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050230450 | description | DESCRIPTION OF THE PREFERRED EMBODIMENT An embodiment of the present invention will now be described with reference to the drawings. FIG. 1 is a block diagram for describing the procedure of plant malfunction diagnosis, FIG. 2 is a diagram showing the overall construction of an advanced thermal reactor (ATR), FIG. 3 is a block diagram showing a hardware configuration for performing plant malfunction diagnosis, FIGS. 4(A), 4(B) and 4(C) are diagrams showing the correlation between frequency spectra of neutron flux and feedwater flow rate in cases where reactivity declines, feed flow rate declines and the pressure of a steam drum increases, respectively, FIG. 5 is a diagram showing a model of a neural network, FIG. 6 is a diagram in which the correlation between frequency spectra is shown as a pattern, FIG. 7 is a diagram showing the learning of connection weights in a neural network, and FIG. 8 is a table showing the results of malfunction cause identification. Shown in these drawings are a calandria tank 1, a control rod 2, a pressure tube assembly 3, a heavy water cooling system 4, a steam drum 5, a coolant recirculating system 6, a water drum 7, a recirculating pump 8, a feedwater system 10, a main steam system 11, an outlet pipe 12, an inlet pipe 13, electronic computers 21, 24, a pattern memory device 22, plants 23, 31, a computer 32, an external memory 33, and an alarm device 34. The principle of the plant malfunction diagnostic method of the present invention will now be described. In carrying out plant malfunction diagnosis, the first step is to obtain the response of the plant parameters after the occurrence of an emergency. Since a malfunction cannot actually be made to occur in the plant in order to determine its plant response at an emergency, a simulation is performed using a simulation code, and the response of plant parameters, at an emergency when an external disturbance is applied is predicted. For example, in a case where a drop in steam drum water level is selected as the type of malfunction, events (a) through (e) indicated below are selected, as inputtable the response of plant parameters after an occurrence of one the events can be obtained from calculation using the simulation codes, from a fault tree obtained by successively connecting, in the form of a tree, cause events each of which results in another event. For example, the above-described event (the drop in steam drum water level) is adopted as the event at the top of the tree, to this is connected a cause event considered to be that which results in the top event, and to this cause event is connected a cause event considered to be that which results in the former cause event, and so on. The events selected are, by way of example: (a) a drop in reactivity; PA1 (b) a drop in feedwater flow rate; PA1 (c) an increase in the pressure of the steam drum; PA1 (d) a drop in feedwater temperature; and PA1 (e) an increase in main steam flow rate. PA1 y.sub.i,o : value of plant state variable under full power operation when type of plant state variable is i; PA1 y.sub.i,t : value of plant state variable at time t when type of plant state variable is i PA1 x.sub.k (t): deviation value at time t of plant state variable of type k PA1 U(t): white noise PA1 M: dimension of autoregressive coefficient PA1 (1) An input signal is fed into the input layer. PA1 (2) A change in the state of each neuron is successively calculated as the signal is transmitted from the input layer to the output layer. PA1 (b 3) In accordance with a back-propagation method, O.sub.i is adopted as the output of the i-th neuron of the output layer obtained from the result of the calculation in (2), and d.sub.i is adopted at the desirable output (an teaching signal) of the neuron corresponding to the input signal. Then, the value of the weight of connection strength is varied so as to minimize the square error ##EQU4## of the difference between the desirable output and the output actually obtained. A learning signal .delta..sub.i.sup.N of the i-th neuron of an N-th stage in and after the middle layer can be determined from recursive calculation in accordance with the following Eq. (7) using a learning signal .delta..sub.k.sup.N+1 of each neuron of an (N+1)th stage: EQU .delta..sub.i.sup.N =f.sub.i.sup.' (u.sub.i.sup.N) .SIGMA. .delta..sub.k.sup.N+1 w.sub.ki.sup.N+1,N (7) PA1 u.sup.N : the internal state of the i-th neuron of the N-th stage ##EQU5## f.sub.i : is the output function of the i-th neuron of the N-th stage EQU (x.sub.i.sup.N =f.sub.i (u.sub.i.sup.N)) PA1 f.sub.i : the differentiated value of f.sub.i PA1 (4) A learning process in which a certain input signal is applied and the desirable output signal corresponding thereto is presented to change the weight of connection strength is performed repeatedly with respect to a set of output instructor signals corresponding to various input signals. The rule for changing the weight of connection strength for every learning process is given by the following Eq. (9) using the learning signal obtained from the procedure of (3): EQU .DELTA.w.sub.ij.sup.N,N-1 (n+1)=.eta..delta..sub.i.sup.N x.sub.j.sup.N-1 +.alpha..DELTA.w.sub.ij.sup.N,N-1 (n) (9) PA1 n: number of learning cycles PA1 .eta.: learning constant PA1 .alpha.: stabilizing constant The occurrence of a drop in steam drum water level corresponds to a malfunction that leads to loss of water which, in an advanced thermal reactor (ATR) of the kind shown in FIG. 2, for example, is used for removing the heat produced by the reactor core, and this in turn can lead to core melt-down. Consequently, the steam drum water level is constantly monitored by the operator very carefully. In the ATR of FIG. 2, heavy water is circulated as a moderator within the calandria tank 1, the heat produced in the moderator is removed by the heavy water cooling system 4, the space between the calandria tubes and the pressure tubes is filled with carbon dioxide gas to serve as a heat insulator, a coolant is circulated through the pressure tubes by the coolant recirculating system 6, water is fed by the feedwater system 10, and steam generated by the steam drum 5 where the feed water flows into is extracted through the main steam system 11 in order to be utilized. Next, with regard to plant state variables obtained from the plant dynamic characteristic analysis, a deviation from steady operation is obtained as shown by the following equation: ##EQU1## where x.sub.i,t : deviation value of plant state variable when type of plant state variable is i and time is t; Eq. (1) is normalized and made dimensionless in such a manner that plant state variables having different dimensions, such as pressure, temperature and flow rate, can be compared with one another. Next, with regard to a plant state variable normalized by computer using autoregressive analytic codes, a pattern indicating the correlation between the frequency spectra of state variables is created. Among the autoregressive analytic codes, time series data X(t) can be expressed as the following equation (2): ##EQU2## where X(t): transposed matrix of [x.sub.1 (t), x.sub.2 (t), . . . x.sub.k (t)] A spectrum density function P(f) can be obtained as indicated by the following equation from the residual covariance matrix .SIGMA. and autoregressive coefficient A(m): EQU P(f)=(A(f)).sup.-1 .SIGMA.((A(f)).sup.T).sup.-1 (3) A(f) is a matrix in which a (j,s) element is A.sub.js (f), P(f) is a matrix in which a (i,j) element is P.sub.ij (f), ().sup.-1 is an inverse matrix, ().sup.T is a transposed matrix and (-) is a complex conjugate. Further, A.sub.js (f) is obtained by a Fourier transformation of the autoregressive coefficient A(m), as indicated by the following equation: ##EQU3## The correlation between the frequency spectra of the plant state variables i and j can be expressed by a coherency function given by the following equation: EQU COH(f).sub.ij =.vertline.P(f).sub.ij .vertline./(P(f).sub.ii .multidot.P(f).sub.jj) (5) Within the ATR of FIG. 2, by way of example, the types of plant state variables for which the correlation between these frequency spectra is examined are neutron flux, recirculation flow rate and steam drum water level in the recirculating system, feedwater flow rate and feedwater temperature in the feedwater system, main steam flow rate, pressure in a steam line and main steam regulative valve opening in the main steam system, and turbine output in the turbine system. The correlations (coherency functions) between the frequency spectra of neutron flux and feedwater flow rate from among these state variables are an indicated in FIG. 4. FIG. 4(A) is for a case where reactivity drops, 4(B) for a case where feedwater flow rate drops, and 4(C) for a case where there is an increase in pressure at the steam drum vapor phase portion. In case of a drop in reactivity, the speed of the malfunction changed at a rate of 2.70.times.10.sup.-2 $/sec, 1.62.times.10.sup.-2 $/sec and 0.41.times.10.sup.-2 $/sec. In the case of feedwater flow rate, flow rate diminished in the form of a step function to 3%, 5% and 7%. In case of an increase in pressure at the steam drum, pressure increased in the form of a step function to 2%, 4% and 6%. In FIGS. 4(A) through 4(C), frequency is plotted along the horizontal axis and the magnitude of the correlation is plotted along the vertical axis. On a broad survey of these Figures, it can be said that the patterns differ from one cause of a malfunction to another without dependence upon the speed or magnitude of the malfunction. When viewed locally, however, these patterns fluctuate when the speed or magnitude of the malfunction is changed. In other words, the correlation value possesses a certain width with respect to a certain single frequency. This width is indicated by applying the net in the Figures. Connection weights in the neural network are learned using these patterns that prevail at a malfunction, patterns obtained with respect to actual plant state variables which change from one minute to the next in the model of the neural network are inputted using the connection weights after learning, and the cause of the malfunction is identified. First, a model of a neural network will be illustrated, then a neural network learning rule, and finally an embodiment of learning. FIG. 5 is a diagram showing a neural network model, in which numeral 41 denotes an input layer, 42 a middle or hidden layer, and 43 an output layer. As shown in the model of the neural network, the network comprises three layers, namely the input layer 41, middle or hidden layer 42 and output layer 43. The number of elements is 96 in the input layer 41, 2-9 in the middle or hidden layer 42, and 5 in the output layer 43. The method in which the elements in the input layer are utilized will be described in detail in the embodiment of learning. The number of elements in the middle or hidden layer is chosen to be between two and nine to minimize computer load. Since the elements in the output layer are made to correspond to the malfunctions to be identified, the number thereof is also made to correspond to the number of malfunctions to be identified. Five malfunctions will be considered here, namely a drop in reactivity, a drop in feedwater flow rate, an increase in pressure at the steam drum, a drop in feedwater temperature, and an increase in main steam flow rate. A learning rule will now be illustrated. Here The initial value of .delta., namely the learning signal .delta..sub.i.sup.o of the output layer, is obtained in accordance with the following Eq. (8): EQU .delta..sub.i.sup.0 =(d.sub.i -O.sub.i)f'(u.sub.i) (8) Accordingly, the values of the learning signals are calculated successively from the output layer toward the input layer using the initial value u.sub.i.sup.o of the internal state u.sub.i.sup.N of the i-th neuron in the output layer, as well as Eq. (7). Input signals and the teaching signals corresponding thereto are presented to the neural network one after another every step. By thus building up the number of learning cycles, the sum of the squared errors indicated by Eq. (6) is reduced. An embodiment of learning will now be illustrated. First, the 96 elements in the input layer are classified, then an example of plant state variables used in identifying the cause of malfunctions is illustrated. Sixty of the elements in the input layer are used to identify input patterns, namely the shapes of the coherency functions. To this end, as shown in FIG. 6, a plane indicating a spectrum is divided into 60 rectangles. A rectangle in which the value of the coherency falls within its limits is expressed by a signal "1" (the shaded portions), and a rectangle in which the value of the coherency does not fall within its limits is expressed by a signal "0" (the white portions). The sixty elements of the input layer bear the respective rectangles as an approximate expression partitioned into 60 portions. When partitioning by the rectangles, initially the horizontal axis of the coherency function from 0 Hz to 0.3 Hz is equally divided into six segments. The vertical axis indicating the correlation values from 0 to 1 is divided into ten portions, but this partitioning is not performed equally. Rather, a range in which the correlation values are small is partitioned roughly, and a range in which the correlation values are large is partitioned finely to raise the analytical precision in case of large correlation values. The range from 0.0 to 0.4 is divided at the 0.2 mark, the range from 0.4 to 0.8 is partitioned into four zones every 0.1, and the range from 0.8 to 1.0 is partitioned into four zones every 0.05. Thirty-six elements are used in order to obtain the type of combination of plant state variables, namely the coherency function of a particular combination of plant state variables. The coherency function of a particular combination of plant state variables is identified by these 36 elements. More specifically, which coherency function of a combination is taught depending upon which element is provided with a "1" signal. The partitioning by 60 is an operation for the purpose of reducing the amount of input information. It is permissible to increase the number of partitions if the processing capability is sufficient. There are five combinations plant state variable used, namely a combination of neutron flux and main steam flow rate, a combination of steam drum water level and main steam flow rate, a combination of recirculation flow rate and main steam flow rate, a combination of pressure at the steam line and main steam regulating valve opening, and a combination of feedwater flow rate and feedwater temperature. Learning is performed until the identification error is reduced to a sufficiently small range of values (e.g., .+-.10%). As for the learning constant, use is made of, e.g., 0.250, which is commonly used in the back-propagation method, and the value 0.9 is employed as the stabilizing constant. Random numeric values are given the initial values of the weight of connection strength and in the order the coherency function are inputted. The method when a pattern collation is carried out will now be illustrated. First, a pattern z.sub.1, z.sub.2, . . . z.sub.n, which indicates the correlation of a frequency spectrum of plant state variables initially created utilizing a simulator is inputted to a neural network, learning is performed until the malfunction cause identification error falls with the desirable range, and the values of the weight of connection strength w.sup.N,N-1, w.sup.N+1,N between the input layer and middle layer and between the middle layer and output layer are decided. This neural network model and the values of the weight of connection strength w are stored as a referential malfunction pattern. When a drop in the steam drum water level is detected, a pattern z.sub.1', z.sub.2', . . . z.sub.n' obtained from the actual plant state variables which change from moment to moment are inserted into the neural network model which uses these weight, and the cause of the malfunction is identified. In order to identify the cause of the malfunction using an ambiguous pattern in which a correlation value possesses a certain width with respect to a certain single frequency, utilizing the neural network is powerful. An example of the results of malfunction cause identification are illustrated in FIG. 8. FIG. 8 illustrates a case in which, when a pattern of a coherency function indicating a drop in reactivity is inputted to an input layer, 0.94.+-.0.01 is obtained from output elements corresponding to the drop in reactivity when the output obtained under ideal conditions is 1, and a maximum of 0.05 is obtained from the other output elements. Similarly, when patterns of coherency functions indicating a drop in feedwater flow rate, an increase in steam drum pressure, an increase in main steam flow rate and a drop in feedwater temperature are inputted, outputs of 0.95.+-.0.01, 96.+-.0.02, 0.96.+-.0.01, and 0.94.+-.0.01 are obtained from the respective corresponding output elements. At this time the outputs from the other output elements are maximum, i.e., 0.06, 0.05, 0.04, 0.05, respectively. Accordingly, the output of a direct-cause element can be obtained at a magnitude which enables it to be sufficiently distinguished from the outputs of the other elements. As a result, it is possible to immediately identify the cause. A procedure for diagnosing plant malfunction based on this principle will now be described with reference to FIG. 1. In FIG. 1, the electronic computer 21 has a function for creating a referential malfunction pattern and creates patterns between abnormal plant state variables. Specifically, plant response analysis at the time of malfunction is performed at step (1) using a simulator, a pattern indicating the correlation of the frequency spectra of the plant state variable is created at step (2), and neural network learning is performed at step (3) to decide the model of the neural network and the connection weight. Next, at step (4), the model of the neural network created at step (3) and the value of the connection weight in the model are stored, as a reference malfunction pattern, in the external memory 22, comprising a magnetic disk, by way of example. At step (5) the computer 24 for plant malfunction diagnosis gathers the plant state variables from the actual plant 23 changing from moment to moment, and at step (6) the computer creates a pattern between plant state variables just as in the case where the pattern was formed at step (2). If the value indicated by a water level indicator of the steam drum is has become smaller than usual, a malfunction is detected at step (7). With regard to the threshold value for determining whether the malfunction has or has not occurred in this case, use is made of a value slightly larger than a value (x.sub.2 .+-..delta.x.sub.2) which fluctuates owing to noise contained in the measurement line (x.sub.2) of the water level indicator. Next, at step (8), the pattern between plant state variables of step (6) is inputted to the neural network model, for which the connection weight has been determined, stored as the referential malfunction pattern, and the cause of the malfunction is identified as described above. The result of the malfunction identification is outputted to a computer printer or the like at step (9). Thus, if the identification of the cause of the malfunction is done by using the neural network, a warning in the control room is issued as by flashing an indicator lamp. If the cause is not clarified, operation returns to the start of plant malfunction diagnosis. In this case, it is permissible to take measures such as presenting a "CONFIRMING OPERATION" indication to the operator. Thus, as set forth above, a cause of malfunction can be identified by utilizing a neural network to be able to distinguish a change in the patterns between plant state variables, such as the neutron flux, temperature, flow rate, pressure and valve opening in a nuclear power plant, which can be readily monitored in a central control room. FIG. 3 is a diagram showing the hardware configuration for performing this malfunction diagnosis. Numeral 31 denotes the plant, 32 the electronic computer, 33 the external memory device and 34 the alarm device, the actions of which are as described above. In accordance with the present invention as described above, the overall state of a plant is judged in a comprehensive manner where collation between changes in plant parameters is investigated. As a result, the cause of a minor malfunction is identified and the operator is informed accordingly, whereby a recovery operation can be carried out at an early stage and it becomes possible to improve not only plant rate of operation but also the safety of the plant. As many apparently widely different embodiments of the present invention can be made without departing from the spirit and scope thereof, it is to be understood that the invention is not limited to the specific embodiments thereof except as defined in the appended claims. |
052079805 | claims | 1. A guide pin assembly for aligning a nuclear fuel assembly with an upper core plate of a nuclear reactor core, said guide pin assembly comprising: (a) an elongated guide pin body having a base portion being insertable within a hole in the top nozzle, said guide pin body having a longitudinal axis; (b) an expandable body insertable within the top nozzle hole with said base portion of said guide pin body, said expandable body having a wall portion including a plurality of circumferentially spaced vertical slots defining between them a plurality of flexible wall segments being capable of expanding radially outwardly relative to said longitudinal axis of said guide pin body to provide an interference fit with the top nozzle; and (c) means insertable within the top nozzle hole with said expandable body and said base portion of said guide pin body for threading with said base portion of said guide pin body to produce a predetermined displacement of said threading means relative to said guide pin body along said longitudinal axis thereof sufficient to impart a radially and outwardly directed force on said flexible wall segments of said wall portion of said expandable body to produce expanding thereof within the hole of the top nozzle into said interference fit with the top nozzle and thereby secure said guide pin body to the top nozzle. (a) an elongated guide pin body having a longitudinal axis, said guide pin body also having a lower expandable base being insertable within a hole i the top nozzle, said lower expandable base including an upper cylindrical base portion, a middle expandable wall portion, and a lower cylindrical skirt portion, said middle expandable wall portion of said lower expandable base having a plurality of circumferentially spaced vertical slots defining between them a plurality of flexible wall segments extending between and interconnecting said upper base portion and said lower skirt portion, said flexible wall segments of said middle expandable wall portion being capable of expanding radially outwardly relative to said longitudinal axis to provide an interference fit with the top nozzle; (b) means insertable within the hole in the top nozzle and interfitted with said guide pin body for imparting a radially and outwardly directed force on said flexible wall segments of said middle expandable wall portion of said lower expandable base to expand said flexible wall segments of said base within the hole of the top nozzle into said interference fit with the top nozzle and thereby secure said guide pin body to the top nozzle in response to a predetermined displacement of said imparting means relative to said guide pin body along said longitudinal axis thereof; and (c) means insertable within the hole in the top nozzle and interfitted with said imparting means and said guide pin body for threading with said guide pin body to produce said predetermined displacement of said imparting means. means defining a hole through said upper base portion of said lower expandable base of said guide pin body in transverse and offset relation to a longitudinal axis of said guide pin body; and a lock pin inserted into said hole in said guide pin body so as to extend past and not intersect with said reduced diameter section of said shank of said threading means and underlie said annular stop on said top end of said shank so as to prevent removal of said lock screw from said guide pin body without first removing said lock pin. means defining a hole through said upper base portion of said lower expandable base of said guide pin body in transverse and offset relation to a longitudinal axis of said guide pin body; and a lock pin inserted into said hole in said guide pin body so as to extend past and not intersect with said reduced diameter section of said shank of said force-imparting means and underlie said annular stop on said top end of said shank so as to prevent removal of said force-imparting means from said guide pin body without first removing said lock pin. (a) an elongated guide pin body having a lower attachment base being insertable within a hole in the top nozzle, said guide pin body having a longitudinal axis; (b) an expandable insert insertable within the hole in the top nozzle and interfitted about said lower attachment base of said guide pin body, said expandable insert including an upper cylindrical base portion, a middle expandable wall portion, and a lower cylindrical skirt portion, said middle expandable wall portion of said expandable insert having a plurality of circumferentially spaced vertical slots defining between them a plurality of flexible wall segments extending between and interconnecting said upper base portion and said lower skirt portion, said flexible wall segments of said middle expandable wall portion being capable of expanding radially outwardly relative to said longitudinal axis of said guide pin body to provide an interference fit with the top nozzle; and (c) means insertable within the hole in the top nozzle and interfitted with said guide pin body and said expandable insert for threading with said guide pin body to produce a predetermined displacement of said threading means relative to said guide pin body along said longitudinal axis thereof sufficient to impart a radially and outwardly directed force on said flexible wall segments of said middle expandable wall portion of said expandable insert to produce expanding thereof within the hole of the top nozzle into said interference fit with the top nozzle and thereby secure said guide pin body to the top nozzle. an end cap attachable on said lower mounting section of said lower attachment base of said guide pin body below said threading means. 2. The guide pin assembly as recited in claim 1, wherein said expandable body is formed integrally with said base portion of said guide pin body. 3. The guide pin assembly as recited in claim 1, wherein said expandable body is removably insertable about said base portion of said guide pin body. 4. A guide pin assembly mountable to a top nozzle of a nuclear fuel assembly for aligning the fuel assembly with an upper core plate of a nuclear reactor core, said guide pin assembly comprising: 5. The guide pin assembly as recited in claim 4, wherein said guide pin body also has an upper end nose and an elongated middle cylindrical shaft portion of generally uniform or constant diameter extending between and integrally connected with said upper end nose and said lower expandable base, said upper end nose and middle shaft portion projecting above the hole in the top nozzle. 6. The guide pin assembly as recited in claim 5, wherein said guide pin body has an external annular flange formed thereon at the juncture of said lower expandable base and said middle shaft portion thereof, said annular flange projecting radially outwardly from said guide pin body and having a downwardly-facing, outwardly and upwardly inclined, shoulder adapted to seat on a complementarily-shaped internal annular surface surrounding the top nozzle hole. 7. The guide pin assembly as recited in claim 4, wherein said upper base portion of said lower expandable base of said guide pin body has a central internally-threaded bore. 8. The guide pin assembly as recited in claim 7, wherein said threading means has a shank with an upper externally-threaded portion adapted to thread within said internally-threaded bore of said lower expandable base of said guide pin body. 9. The guide pin assembly as recited in claim 8, wherein said threading means also has an annular stop on a top end of said shank and a reduced diameter section between said stop and said upper threaded portion of said shank, said reduced diameter section and stop adapted to fit upwardly into said threaded bore in said upper base portion of said lower expandable base of said guide pin body. 10. The guide pin assembly as recited in claim 9, further comprising: 11. The guide pin assembly as recited in claim 4, wherein said wall segments of said middle expandable wall portion have respective interior surfaces being inclined upwardly and inwardly with respect to a longitudinal axis of said guide pin body. 12. The guide pin assembly as recited in claim 11, wherein said force-imparting means has an exterior surface inclined upwardly and inwardly relative to the longitudinal axis of said guide pin body and thus complementary to and interfaced with said interior surfaces of said wall segments of said lower expandable base. 13. The guide pin assembly as recited in claim 4, wherein said force-imparting means has a central opening. 14. The guide pin assembly as recited in claim 13, wherein said tightening means has a shank fitting upwardly through said central opening of said force-imparting means. 15. The guide pin assembly as recited in claim 14, wherein said force-imparting means has a lower head integrally attached to a lower end of said shank. 16. The guide pin assembly as recited in claim 15, wherein said lower skirt portion of said lower expandable base of said guide pin body has a lower peripheral edge crimped under an outer peripheral edge of said lower head of said force-imparting means. 17. The guide pin assembly as recited in claim 16, wherein said lower head of said force-imparting means has a plurality of relief pockets formed in spaced circumferential relation from one another about said peripheral edge of said head, said lower skirt portion being crimped into at least some of said relief pockets. 18. The guide pin assembly as recited in claim 16, wherein said lower expandable base of said guide pin body has a central internally-threaded bore. 19. The guide pin assembly as recited in claim 18, wherein said shank of said force-imparting means has an upper externally-threaded portion adapted to thread within said internally-threaded bore of said lower expandable base of said guide pin body. 20. The guide pin assembly as recited in claim 19, wherein said force-imparting means also has an annular stop on a top end of said shank and a reduced diameter section between said stop and said upper threaded portion of said shank, said reduced diameter section and stop adapted to fit upwardly into said threaded bore in said upper base portion of said lower expandable base of said guide pin body. 21. The guide pin assembly as recited in claim 20, further comprising: 22. A guide pin assembly for aligning a nuclear fuel assembly with an upper core plate of a nuclear reactor core, said guide pin assembly comprising: 23. The guide pin assembly as recited in claim 22, wherein said guide pin body also has an upper end nose and an elongated middle cylindrical shaft portion of generally uniform or constant diameter extending between and integrally connected with said upper end nose and said lower attachment base, said upper end nose and middle shaft portion projecting above the hole in the top nozzle. 24. The guide pin assembly as recited in claim 23, wherein said guide pin body also has a locking disc attached thereto at the juncture of said middle shaft portion and said lower attachment base. 25. The guide pin assembly as recited in claim 24, wherein said expandable insert has an upper base portion with a plurality of relief pockets provided in a top surface of said upper base portion, said locking disk being crimpable into at least some of said relief pockets to prevent rotation of said guide pin body relative to said expandable insert. 26. The guide pin assembly as recited in claim 22, wherein said expandable insert includes an annular flange formed about an upper peripheral edge of said upper base portion thereof, said flange projecting radially outwardly from said upper base portion and having a downwardly-facing, outwardly and upwardly inclined, shoulder adapted to seat on a complementarily-shaped internal annular surface defined about the top nozzle hole. 27. The guide pin assembly as recited in claim 22, wherein said lower attachment base includes has upper and lower cylindrical mounting sections and an externally-threaded middle mounting section. 28. The guide pin assembly as recited in claim 27, wherein said upper base portion of said expandable insert has a central bore receiving said upper mounting section of said lower attachment base of said guide pin body. 29. The guide pin assembly as recited in claim 22, wherein said wall segments of said expandable wall portion have respective interior surfaces being inclined upwardly and inwardly with respect to said longitudinal axis of said guide pin body. 30. The guide pin assembly as recited in claim 29, wherein said threading means has an exterior surface inclined upwardly and inwardly relative to said longitudinal axis of said guide pin body and thus complementary to and interfaced with said interior surfaces of said wall segments of said expandable wall portion of said expandable insert. 31. The guide pin assembly as recited in claim 27, wherein said threading means has an internally-threaded central opening for threading said threading means onto said externally-threaded middle section of said lower attachment base of said guide pin body. 32. The guide pin assembly as recited in claim 31, further comprising: 33. The guide pin assembly as recited in claim 32, wherein said end cap is disposable within said lower skirt portion of said expandable insert. 34. The guide pin assembly as recited in claim 22, wherein said threading means also includes a plurality of radially outward projecting tabs formed on said exterior surface thereof, said tabs being spaced circumferentially from one another so as to project into said slots in said expandable middle wall portion of said expandable insert so as to prevent rotation of said threading means relative to said expandable insert as said threading means is threaded with said guide pin body. |
summary | ||
abstract | The invention relates to a nuclear power plant including a containment vessel including a reactor pressure vessel for receiving fissionable nuclear fuel, an aerosol filter stage a pressure relief conduit through which a gas volume flow which is filtered in the aerosol filter stage is releasable to ambient through a pass through opening in the containment vessel, and an iodine filter stage through which the gas volume flow that is filtered in the aerosol filter stage is filterable before being released to the ambient, wherein the iodine filter stage is arranged within the containment vessel, characterized in that the aerosol filter stage and the iodine filter stage are connected with one another so that transferring the gas volume flow from the aerosol filter stage to the iodine filter stage is performed essentially at an identical pressure level. |
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055132308 | abstract | A fuel rod for a nuclear reactor includes a metal cladding tube filled with nuclear fuel and having ends, an outer surface, and a longitudinal axis. A metal seal plug is welded to one of the ends of the cladding tube at a transition point, defining an annular bead on the outer surface of the cladding tube at the transition point. The annular bead has a cylindrical outer jacket surface with jacket lines being substantially parallel to the longitudinal axis of the cladding tube. The annular bead has material being formed of the metal of the cladding tube and the metal of the seal plug. A welding apparatus for producing a fuel rod includes an electrode having a through bore formed therein for receiving one end of a cladding tube. A counter electrode is displaceable relative to the electrode for holding a seal plug to be welded the one end of the cladding tube. The electrode has a cylindrical step formed therein at an end of the through bore facing toward the counter electrode. The cylindrical step has a diameter being greater than the diameter of the through bore. |
abstract | An X-ray image photographing apparatus includes an image obtaining portion for obtaining an X-ray distribution transmitted through an object, a grid detecting system having a construction for obtaining information from a grid side by the action of inserting the grid for decreasing scattered rays into the apparatus, and detecting at least one of the presence or absence of the grid, the kind of the grid and the presence or absence of the replacement of the grid by the use of the construction, an image processing system for image processing and outputting image data collected by the image obtaining portion, and a memory portion preserving therein a plurality of sets of image processing parameters for controlling the image processing system. The image processing system selects the image processing parameters preserved in the memory on the basis of at least the result of the detection by the grid detecting system and executes image processing. |
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abstract | The spacer/water rod retention assembly includes a clip secured to the spacer having a slot perimetrically bounded both axially and laterally spaced portions of the clip. A resilient tab projects from one side portion into the slot. A water rod has a projecting tab which, upon relative rotation of the spacer and water rod, engages and deflects the clip tab to enable the water rod tab to reside in the clip slot. When the clip tab returns to its initial position, substantial relative rotation between the water rod and spacer is precluded and axial connection between the spacer and water rod is assured. In another form, the water rod tab projects into a slot of the clip. An adjacent water rod prevents rotation of the water rod tab in the opposite direction, thereby capturing the spacer and water rod and preventing relative axial movement. |
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047568771 | description | DETAILED DESCRIPTION Referring now to the drawings, FIG. 1 illustrates the lower portion of a nuclear reactor 1 which has a substantially cylindrical, vertically oriented pressure vessel 3, the pressure vessel having a vertical wall 5 and hemispherical bottom wall 7. A core barrel 9 is disposed within the pressure vessel 3, the core barrel 9 having a bottom core support plate 11, which supports various components of the reactor constituting the core internals such as the fuel element assemblies, the fuel element assemblies support grid structure, and the like, designated at 13, all as conventionally provided. The upper section of the core barrel 9 conventionally is provided with an annular, radially outwardly extending flange, not shown, and an upper region of the pressure vessel 3 has an annular, radially inwardly extending ledge, also not shown, upon which the core barrel flange is seated. Spaced about the upper wall portion 15 of the hemispherical bottom wall 7 are a plurality of engagement means 17, four such engagement means being shown in the drawings, although the number thereof may vary dependent upon the size and type of reactor. The engagement means 17 preferably have a vertical flow channel therethrough and is also preferably formed as a pair of brackets 19 that are attached to the inner wall surface 21 of the upper wall portion 15 of the hemispherical bottom wall 7, and extend radially inwardly therefrom, and a cross-bar or connecting portion 23 having a recess 25 therein, with a flow channel 27 formed between the wall surface 21, brackets 19 and connecting portion 23. The core support plate 11 is thus engaged about its periphery by a plurality of engagement means. In accordance with the present invention, the core support plate 11 is aligned and secured on the engagement means 17 within the pressure vessel 3 without the need for the entry of personnel into the area 29 between the core support plate 11 and the hemispherical bottom wall 7 of the pressure vessel. The core support plate 11 is provided with a plurality of apertures 31, preferably having a circular wall, which extend vertically therethrough, about the peripheral region thereof, also preferably with a countersunk portion 33 adjacent the upper surface 35 of the core support plate 11 which forms an annular shoulder 37. An aperture 31 is provided in the core support plate 11 for alignment with each of the engagement means 17, above the recess 25 of the engagement means. To align and affix the core support plate 11 to the engagement means 17, key members 39 are provided, an embodiment of which is illustrated in FIGS. 4 and 5. With the use of apertures 31 having a circular wall thereabout, the key member has a cylindrical upper section 41, a lower non-cylindrical, and preferably rectangular section 43, and a top outwardly radial flange section 45, the upper surface 47 of which is flat. As illustrated, the keys 39 are secured in the apertures 31, with the flanged section 45 fitting within the countersunk portion 33 of the core support plate 11, and the top surface 47 of the key is flush with the top surface 35 of the core support plate 11. The keys are then secured to the core support plate such as by welding or by the use of bolts 49 which extend through bores 51 in the flange 45 that engage in complementary bores 53 in the shoulder 37 formed in the core support plate 11 about the aperture 31. Upon securement, the lower section 43 of the keys 39 extend into and register in the recesses 25 in the engagement means 17 to provide lateral and radial support and alignment. The lower section 43 of the key 39 is machined so as to provide a mating relationship in the recess 25 of the engagement means, or, if desired, shims 55 may be added to provide such relationship. The provision of collapsible shock absorbers between the core support plate 11 and the engagement means 17 is provided by placement of collapsible cylinders such as stainless steel cylinders, 57 in cavities 59 formed in the surface 61 forming recess 25 and in complementary cavities 63 formed in the bottom wall 65 of the key 39, the cylinders preferably of a mushroom shape. In the method of installing and aligning a core barrel in a nuclear reactor pressure vessel according to the present invention, a pressure vessel 3 is provided with a plurality of engagement means 17, having recesses 25 therein, about the lower inner wall thereof, i.e., spaced about the inner wall surface 21 of the upper wall portion 15 of the hemispherical bottom wall 7 of the pressure vessel 3. A core support plate 11 of a core barrel 9 is provided with a plurality of apertures 31 therethrough, with an aperture 31 provided for alignment with each of the recesses 25 of the engagement means 17. The core barrrel 9 with its core support plate 11 is inserted into the pressure vessel 3 from the top and the apertures 31 aligned with the recesses 25. The assembler then determines, through the apertures 31, the exact dimensions required for the lower section 43 of each key 39 and the key is machined to the desired tolerances. Where desired, the provision of shims to provide the desired clearance may also be manufactured and inserted. The keys 39 are then inserted from the top, into the apertures 31 with the lower section 43 fitted in the recesses 25. The keys 39 are then secured to the top of the core support plate 11 by welding or by use of bolts 49 with the upper surface 47 of the key 39 being flush with the upper surface 35 of the core support plate 11. The present invention permits the measurement of the relative position of the keys in the core support plate and the recesses or keyways in the engagement means without physically locating an assembler in the area between the core support plate and the hemispherical bottom wall of the pressure vessel. With the present arrangement, the conventional manway and cover plug of the core support plate can be eliminated. Also, with the present invention, the need to remove the core barrel from the reactor vessel for insertion of custom-made shims in the recesses of the engagement means is not required. |
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abstract | A cooling system for a reactor module includes a reactor pressure vessel that houses primary coolant and a steam generator that lowers a temperature of the reactor pressure vessel by transferring heat from the primary coolant to a secondary coolant. The steam generator releases at least a portion of the secondary coolant as steam. Additionally, the cooling system includes a containment vessel that at least partially surrounds the reactor vessel in a containment region. The containment region is dry during normal operation of the reactor module. A controller introduces a source of water into the containment region in response to a non-emergency shut down of the reactor module. The source of water is located external to the containment vessel, and the water is introduced into the containment region after the steam generator has initially lowered the temperature of the reactor pressure vessel in response to releasing the steam. |
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claims | 1. A device for producing a fluid containing a radioactive constituent, the device comprising: a shielded chamber within which is located an isotope container housing a radioactive isotope, the shielded chamber including first and second fluid connections to opposing ends of the isotope container and a fluid conduit extending from each of the first and second fluid connections to a fluid inlet and a fluid outlet respectively characterised in that the fluid inlet comprises a single spike comprising a spike body having a substantially circular cross-section, the spike body being adapted to penetrate the rubber seal of a vial and wherein the spike body further defines a first bore extending from a first aperture adjacent the tip of the spike to a fluid connection with the fluid conduit and a second bore extending from a second, separate aperture in the spike body to a filtering air inlet, wherein the spike body incorporates a barrier filter in the second bore. 2. A device as claimed in claim 1 further comprising an outer housing which supports the fluid inlet and the fluid outlet and the spike of the fluid inlet projects through an aperture in the outer housing. 3. A device as claimed in claim 2, wherein the outer housing defines a well about the aperture through which the spike projects, the well being structured to receive a vial. 4. A device as claimed in claim 1, wherein the barrier filter comprises a filter disk of polytetrafluoroethylene. 5. A device as claimed in claim 1, wherein the filter is positioned in the second bore adjacent to the filtering air inlet. |
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043127070 | abstract | A nuclear fuel rod, which comprises:. a cladding tube filled with an inert gas or liquid metal and sealed at both ends in airtightness; and PA0 a plurality of nuclear fuel pellets piled one atop another in the cladding tube,. and wherein an adsorbent carrying a tag gas for monitoring a nuclear fuel rod failure is received in the inner space of the cladding tube defined at least above or below the nuclear fuel pellet pile. |
claims | 1. A method of providing electrical system monitoring and diagnosis, comprising:providing a motor controller including solid state switches for controlling application of power to the motor, and a control circuit for controlling operation of the solid state switches and for measuring electrical power system characteristics relating to operation of the solid state switches;providing an external monitoring and diagnostic device;establishing communications between the control circuit and the external monitoring and diagnostic device; andperiodically transferring parameters of the measured electrical power system characteristics from the control circuit to the external monitoring and diagnostic device to monitor electrical power system characteristics in real time. 2. The method of providing electrical system monitoring and diagnosis of claim 1 wherein providing a motor controller comprises providing a control circuit including a programmed processor for commanding operation of the solid state switches and a memory connected to the programmed processor for storing the parameters of the measured electrical power system characteristics. 3. The method of providing electrical system monitoring and diagnosis of claim 2 wherein transferring parameters of the measured electrical power system characteristics comprises reading the stored parameters of the measured electrical power system characteristics from the memory. 4. The method of providing electrical system monitoring and diagnosis device of claim 1 wherein providing an external monitoring and diagnostic device comprises providing a computer having a memory for storing the transferred parameters. 5. The method of providing electrical system monitoring and diagnosis of claim 1 wherein providing an external monitoring and diagnostic device comprises providing a personal digital assistant having a memory for storing the transferred parameters. 6. The method of providing electrical system monitoring and diagnosis of claim 1 further comprising printing a listing of the transferred parameters of the measured electrical power system characteristics. 7. The method of providing electrical system monitoring and diagnosis of claim 1 wherein periodically transferring parameters of the measured electrical power system characteristics comprises transferring the parameters at select time intervals. 8. The method of providing electrical system monitoring and diagnosis of claim 1 wherein the control circuit measures line voltage, motor voltage and motor current. 9. The method of providing electrical system monitoring and diagnosis of claim 1 wherein establishing communications between the control circuit and the external monitoring and diagnostic device comprises providing an infrared communication path between the control circuit and the external monitoring and diagnostic device. 10. The method of providing electrical system monitoring and diagnosis of claim 1 wherein establishing communications between the control circuit and the external monitoring and diagnostic device comprises providing a wired communication path between the control circuit and the external monitoring and diagnostic device. 11. A motor controller system for monitoring and diagnosing electrical power system characteristics, comprising:a motor controller including solid state switches for controlling application of power to a motor, and a control circuit for controlling operation of the solid state switches and for measuring electrical power system characteristics relating to operation of the solid state switches;an external monitoring and diagnostic device including a memory for storing parameters of the measured electrical power system characteristics and an interface for communication with the motor controller; andmeans operatively associated with the control circuit and the external monitoring and diagnostic device for transferring parameters of the measured electrical power system characteristics from the control circuit to the external monitoring and diagnostic device to monitor electrical power system characteristics in real time. 12. The motor controller system of claim 11 wherein the control circuit comprises a programmed processor for commanding operation of the solid state switches and a memory connected to the programmed processor for storing the parameters of the measured electrical power system characteristics. 13. The motor controller system of claim 12 wherein the transferring means comprises means for reading the stored parameters of the measured electrical power system characteristics from the memory. 14. The motor controller system of claim 11 wherein the external monitoring and diagnostic device comprises a computer having a memory for storing the transferred parameters. 15. The motor controller system of claim 11 wherein the external monitoring and diagnostic device comprises a personal digital assistant having a memory for storing the transferred parameters. 16. The motor controller system of claim 11 further comprising printer operatively associated with the external monitoring and diagnostic device for printing a listing of the transferred parameters of the measured electrical power system characteristics. 17. The motor controller system of claim 11 wherein the transferring means transfers the parameters at select time intervals. 18. The motor controller system of claim 11 wherein the control circuit measures line voltage, motor voltage and motor current. 19. The motor controller system of claim 11 wherein the transferring means comprises an infrared communication path between the control circuit and the external monitoring and diagnostic device. 20. The motor controller system of claim 11 wherein the transferring means comprises a wired communication path between the control circuit and the external monitoring and diagnostic device. 21. A soft starter system for monitoring and diagnosing electrical power system characteristics, comprising:a motor controller including solid state switches for controlling application of power to a motor, and a control circuit for controlling operation of the solid state switches, the control circuit comprising a programmed processor for commanding operation of the solid state switches and for measuring electrical power system characteristics relating to operation of the solid state switches, and a memory connected to the programmed processor storing parameters of the measured electrical power system characteristics;an external monitoring and diagnostic device including a memory for storing parameters of the measured electrical power system characteristics and an interface for communication with the motor controller; anda monitoring and diagnostic program operatively implemented in the external monitoring and diagnostic device for transferring parameters of the measured electrical power system characteristics from the control circuit to the external monitoring and diagnostic device to monitor electrical power system characteristics in real time. 22. The soft starter system of claim 21 wherein the external monitoring and diagnostic device comprises a computer having a memory for storing the transferred parameters. 23. The soft starter system of claim 21 wherein the external monitoring and diagnostic device comprises a personal digital assistant having a memory for storing the transferred parameters. 24. The soft starter system of claim 21 wherein the monitoring and diagnostic program is operable to upload the parameters from the controller memory to the external monitoring and diagnostic device memory. 25. The soft starter system of claim 21 wherein the control circuit measures line voltage, motor voltage and motor current. 26. The soft starter system of claim 21 wherein the interface comprises a wireless interface. 27. The soft starter system of claim 26 wherein the interface comprises a wired interface. |
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abstract | A voloxidizer with a double reactor for spent fuel rods decladding of the present invention includes a reactor module into which spent fuel rods are loaded, the reactor module including a reactor having a dual structure; a heater module for heating the reactor module; and a drive module for providing a driving force to the reactor module. A double reactor utilized in a voloxidizer for spent fuel rods decladding includes an internal reactor into which spent fuel rods are loaded; and an external reactor formed on an outer circumferential surface of the internal reactor. Here, a first transport part and a second transport part are formed on inside surfaces of the internal reactor and the external reactor, respectively, and the spent fuel rods are moved by the first transport part and the second transport part and oxidized when the internal reactor and the external reactor are rotated. |
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048328987 | description | DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENT FIG. 2 illustrates a simplified schematic representation of a typical pressurized water reactor in which the method and apparatus of the present invention to variably delay steam generator low water level reactor protective functions can be employed. The reactor vessel 20 has coolant flow inlet means 21 and coolant flow outlet means 22. The vessel 20 contains a nuclear core (not shown) consisting mainly of a plurality of clad nuclear fuel elements which generate substantial amounts of heat, depending primarily upon the position of control rods 23. The heat generated by the reactor core is conveyed from the core by coolant flow entering through inlet means 21 and exiting through outlet means 22. The flow exiting through outlet means 22 is conveyed through an outlet conduit 24 to a heat exchange steam generator system 25. The heated coolant is conveyed through heat exchange tubes 26 which are in a heat exchange relationship with water 27 which is used to produce steam. The steam produced by the generator 25 is utilized to drive a turbine 28 for the production of electricity as described more fully below. The flow of the coolant is then conveyed from the steam generator 25 through an inlet conduit 29 to inlet means 21. Thus, a closed recycling primary loop couples the reactor vessel 20 and the steam generator 25. The system shown in FIG. 2 is illustrated with one closed fluid flow loop although the number of loops and hence the number of steam generators varies from plant to plant and commonly two, three, or four are employed. The secondary side of the steam generator 25 is isolated from the primary loop by the heat exchange tubes 26. The water 27 in the steam generator 25 is placed into a heat exchange relationship with the primary coolant, whereby the water 27 is heated and converted to a vapor or steam. The vapor flows through a steam conduit 30 to the turbine 28. The steam, after passing through the turbine 28, is condensed in a condenser 31. The condensate or water is returned to the secondary side of the steam generator 25 through conduit 32. Thus, a recycling, secondary loop couples the steam generator 25 to the turbine 28. Completing the description of the system shown in FIG. 2, a pressure differential sensor 33 measures the pressure differential between pressure taps 34 and 35 and produces a signal 37 representative of the water level 27 in steam generator 25. A sensor 36 measures the neutron flux in the reactor core (not shown) and produces a signal 38 representative of the reactor thermal power output level. FIG. 3 illustrates a simplified schematic of the auxiliary feedwater system in a typical four steam generator plant. Auxiliary feedwater pumps 41, 42, and 43 provide feedwater to steam generators 25 from a condensate storage tank 40, which is the source of the auxiliary feedwater, through auxiliary feedwater conduits 44. Auxiliary feedwater pumps 41 and 43 are driven by motors 45 while auxiliary feedwater pump 42 is powered by a turbine 46. Flow control valves 47 permit the regulation of auxiliary feedwater flow to each steam generator 25. Turning to FIG. 4, the diagram illustrates a steam generator low water level reactor protection system 50 constructed according to the teachings of the present invention. The reactor power output level signal 38 is input to two reactor power output level bistables 74 and 75 which produce output signals 84 and 85, respectively. The number of power output level bistables is indicative of the desired accuracy to which the power level must be ascertained for any particular implementation of this system. The steam generator water level signal 37 may be input in parallel to four separate setpoint bistables 70, 71, 72 and 73 which are provided for each steam generator loop to satisfy safety redundancy requirements. The four water level setpoint bistables 70, 71, 72 and 73 produce signals designated by numerals 80, 81, 82 and 83, respectively. The output signals 80, 81, 82 and 83 are inputs to an AND gate 76. The number of AND gates 76 corresponds to the number of steam generators. In order to insure that a valid low water level condition exists in the steam generator loop, the output signal 86 of the AND gate 76 will only indicate a low water condition if two out of the four water level bistables 70, 71, 72 and 73 indicate such a condition. The output signal 86 from each AND gate 76 drives both an OR gate 77 and an AND gate 78. A low water condition in any steam generator loop will cause a signal 87 to be available at an output of OR gate 77. Similarly, a low water condition in two or more steam generators will cause a signal 88 to be available at an output of AND gate 78. Signals 84 and 85, the output signals produced by power output level bistables 74 and 75, respectively, may correspond to a high logic level when the power output of the reactor is less than fifty percent of Rated Thermal Power in the case of signal 84 and less than ten percent of Rated Thermal Power in the case of signal 85. Conversely, NOT gates 91 and 92, which are driven by signals 84 and 85, respectively, may provide high logic level output signals 93 and 94, respectively, when the power output of the reactor exceeds fifty percent of Rated Thermal Power in the case or NOT gate 91 and ten percent of Rated Thermal Power in the case of NOT gate 92. One NOT gate is provided for each power output level bistable included in the system. Timers 60, 61, 62 and 63 are activated by the signal 87 generated by OR gate 77. Each of these timers is set to time out at a predetermined period of time which is acceptable under certain power level conditions and steam generator low water level conditions. Extensive analyses were performed to insure that all plant operating parameters remained within acceptable levels during the delay periods. Those analyses resulted in the curves shown in FIG. 5. One curve illustrates reactor trip and auxiliary feedwater actuation delay times in thousands of seconds as a function of power level in percent of Rated Thermal Power for a low water condition in one steam generator. The second curve illustrates the same information for a low water condition in two or more steam generators. Referencing these curves, and assuming worst case power output levels, timer 60 may be set for a delay of one hundred fifty seconds (one steam generator low level, power output at fifty percent of Rated Thermal Power), timer 61 may be set for a delay of thirty seconds (one steam generator low level, power output at one hundred percent of Rated Thermal Power), timer 62 may be set at two hundred eighty seconds (two or more steam generators low level, power output at ten percent of Rated Thermal Power) and timer 63 may be set for five seconds (two or more steam generators low level, power output at one hundred percent of Rated Thermal Power). The output signal 66 of timer 60, the output signal 94 of NOT gate 92 and signal 84 drive AND gate 95; the output signal 67 of timer 61 and the output signal 93 of NOT gate 91 drive AND gate 96; the output signal 68 of timer 62 and signal 85 drive AND gate 97; the output signal 69 of timer 63, the output signal 94 of NOT gate 92 and the output signal 88 of AND gate 78 drive AND gate 98. This configuration results in the selection of one or more of timer output signals 66, 67, 68 and 69 the selection of which is safely allowable based upon the reactor power output level and the severity of the steam generator low water level condition. In the event that more than one of timer output signals 66, 67, 68 and 69 is selected, the signal 66, 67, 68 or 69 associated with the timer 60, 61, 62 or 63 corresponding to the shortest delay period will control the reactor protective functions. AND gate 95 may produce a high logic level output at the expiration of one hundred fifty seconds if a low water level condition exists in one steam generator and the reactor power output level is between ten and fifty percent of Rated Thermal Power. AND gate 96 may produce a high logic level output at the expiration of thirty seconds if a low water level condition exists in one steam generator and the power output level of the reactor exceeds fifty percent of Rated Thermal Power. AND gate 98 may produce a high logic level output at the expiration of five seconds if a low water condition exists in two or more steam generators and the reactor power output level is in excess of ten percent of Rated Thermal Power. AND gate 97 may produce a high logic level output at the expiration of two hundred eighty seconds if the reactor power output is less than ten percent of Rated Thermal Power and a low water condition exists in one or more than one steam generator. Timer 62 controls the protective functions in this case regardless of the number of steam generators affected because the output of timer 63 will be blocked by AND gate 98 when the reactor power output level is less than ten percent of Rated Thermal Power. This design permits the selection of different time delays as either the number of steam generators affected by a low water condition changes or the reactor power output level fluctuates during a steam generator low water level condition. Additionally, should any protective action initiating conditions cease to exist, the system is reset and no protective actions taken. System reset is not desired when the reactor power output level has exceeded and then falls below one of the predefined power output levels and, therefore, latches (not shown) are provided. Specifically, the power output signals 84 and 85, and, therefore, signals 93 and 94, are latched so that if power levels fluctuate, reactor trip will still occur unless steam generator water levels return to a normal condition. A signal generated at the output terminal of one of AND gates 95, 96, 97 and 98 will initiate reactor protective functions. A reactor trip signal 99 causes the control rods 23 in FIG. 2 to be inserted into the nuclear core (not shown) and thus halt the generation of heat. An auxiliary feedwater actuation signal 100 actuates motors 45 and turbine 46 in FIG. 3 to provide power to auxiliary feedwater pumps 41, 42 and 43. The auxiliary feedwater actuation signal 100 also opens the flow control valves 47 in FIG. 3 to permit the flow of auxiliary feedwater to steam generators 25. The logic circuit shown in FIG. 4 illustrates one possible configuration of a system to variably delay reactor trip and auxiliary feedwater system actuation. The circuit can be expanded to accommodate any number of steam generator loops. The number of timers can also be varied depending upon how closely the curves of FIG. 5 are to be modeled. An alternate implementation of the steam generator low water level reactor protection system of the present invention involves the use of a microprocessor 110 as shown in FIG. 6 in lieu of the logic circuitry of FIG. 4. As in FIG. 4, the output signal 86 from each AND gate 76 drives both the OR gate 77 and the AND gate 78. The signal 87, available at the output of OR gate 77 when a low water condition exists in any one steam generator, as well as signal 88, available at the output of AND gate 78 when a low water condition exists in two or more steam generators, are inputs to the microprocessor 110. The microprocessor 110 also receives the reactor power output level signal 38. The microprocessor 110 could be programmed with the curves shown in FIG. 5 to determine the appropriate delay, on a real-time basis, based upon the number of steam generators affected by the low water level condition and the reactor power output level. Signals 99 and 100 could then be generated by the microprocessor 110 and output through a known output interface 114 to initiate reactor trip and auxiliary feedwater system actuation, respectively, if the steam generator water level is not restored. A variation on this microprocessor implementation would incorporate a table containing the information of FIG. 5. The microprocessor 110 would be programmed such that the appropriate delay would be determined by referencing the table of delay times based upon the severity of the low water transient and the measured power output level. While the present invention has been described in connection with exemplary embodiments thereof, it will be understood that many modifications and variations will be readily apparent to those of ordinary skill in the art. This disclosure and the following claims are intended to cover all such modifications and variations. |
abstract | A process of treating hydrogen gas liberated from the acid or alkaline dissolution of a metal is provided. The process comprises a step of passing the liberated hydrogen gas through a reactor containing an oxidizing agent for oxidation of the hydrogen gas into water, followed by a step of regenerating the oxidizing agent. Also provided is an apparatus for carrying out the process, the apparatus comprising a reactor containing the oxidizing agent, wherein the reactor is at least partially immersed in an alumina bath. |
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041586816 | description | DETAILED DESCRIPTION OF THE INVENTION The nuclear fuel pellets first travel through a reduction furnace with an adjustable velocity or dwelling time for adjusting the stoichiometry of the nuclear fuel oxides. The pellets are then taken in cooled-down condition to a checking station and pass subsequently through a sintering furnace with independently adjustable dwelling time or throughput velocity. The reduction of such pressed blanks initially in stoichiometric excess of oxygen is accomplished, for example, in an externally heated muffle furnace, the temperature profile of which can be fitted by appropriately controlling individual heater circuits within wide ranges (100.degree. to 1000.degree. C.) to the temperature curve which is optimum for conducting the reaction as a function of various conditions (stoichiometric excess of oxygen, quantity of pressed blanks per unit time, shape of the pressed blanks, reactivity of the powders). In order to keep the water concentration low in accordance with the course of the reduction, additional gas injection stations are provided at different points. After leaving the muffle, the blanks can quickly be checked in the control station to determine whether the reduction has taken place completely. The check can be made, for example, optically; thus, the color hue of the pressed blanks gives information regarding the stoichiometric condition. This has the advantage that it can be ascertained at an early stage of the operation whether the pressed blanks are sufficiently reduced. For example, in case of insufficient reduction of the blanks, defects would be produced in the sintering which would limit their usability. Such inadequately reduced pressed blanks might then even have to be scrapped as rejects. A nitrogen/hydrogen mixture can be used as the reduction gas, which has considerable cost advantages over the use of a mixture of rare gas and hydrogen. A mixture of nitrogen and carbon monoxide may also be used as the reduction gas. Larger amounts of carbon monoxide as compared to hydrogen may be safely tolerated in the gas mixture. The hydrogen content in the gas mixture is usually 4 to 8 % while the carbon monoxide content is 4 to 12 %. The push-through velocity and the dimensions of the muffle can be matched optimally to the given reduction and material conditions. In particular, the push-through velocity is independent of the push-through velocity of the sintering furnace. The extent of reduction of the pellets depends on a number of factors, including throughput or quantity of pellets fed to the furnace per unit time, velocity of the pellets through the furnace, length of the furnace, furnace temperature and gas composition. The effect of these factors on the pellets in the furnace could be referred to as exposure of the pellets to conditions in the furnace, which shall be designated as "residence time". Normally the dimensions of a furnace are fixed by previous design of the furnace. Also to some extent temperature variations are limited by previous design of the furnace. While gas compositions may be varied, in general, since the hydrogen content is not to materially exceed 8 %, changes in the hydrogen content would be small. Usually in an operation, the temperature and gas composition are set and any variation or adjustment or regulation to obtain the desired reduction and sintering of the pellet is readily and conveniently accomplished by varying the velocity of the pellets through the furnace or by varying the throughput, i.e., increasing or decreasing the quantity of pellet passing through the furnace, or a combination of a change of velocity and a change of throughput. Through the use of a closed, externally heated muffle, further advantages accrue due to the fact that the replacement of defective heater circuits is possible without the aggravating conditions of working with radioactively contaminated work pieces; that the gas flow conditions and the mass balance can be controlled better than in furnaces equipped with gas-permeable linings; and finally, that the decomposition products of the lubricant or/and the binder and lubricating agents are no longer condensed uncontrollably at colder points of the furnace jacket but are removed from the muffle with the hot gas. Through using nitrogen/hydrogen mixtures, the employment of electrostatic dust separators becomes furthermore possible, which remove, in addition to dust, also the decomposition products of the lubricating oil and/or of the binder and lubricating agents. Electrostatic separators cannot be used in the presence of argon (the cheapest rare gas), as argon is ionized already at the relatively high voltage and the separator then breaks down. A high degree of separation is necessary, however, as the furnace gas must be discharged to the outside air only via absolute filters. The sintering of the reduced pressing blanks is performed, after they pass through an intermediate or checking station, in a resistance-heated furnace lined with highly refractory blocks. As all or substantially all the reduction had previously been effected in the reduction furnace and no further reduction need take place, the reduction potential in the furnace can be adjusted to any required order of magnitude without effect on the preceding reduction. Overall, only a small quantity of a rare gas/hydrogen mixture is necessary for this purpose, as the material to be sintered is already reduced and no additional water is therefore generated. This is accompanied, in addition, by considerable cost advantages over the process technique customary heretofore. By decoupling the reduction from the sintering, the length of the sintering furnace and the push-through velocity can be optimally matched to the operational requirements such as space required and maximum loading on the one hand and the requirements as to the sintered oxide such as, for example, a minimum dwelling time in the high-temperature zone. The operation is best conducted with temperatures lower in the reduction furnace than in the sintering furnace. In general, the reduction furnace operates from about 700.degree. C. to about 1000.degree. C., preferably about 1000.degree. C., and the sintering furnace operates at from about 1000.degree. C. to about 1760.degree. C., preferably about 1600.degree. C. to 1700.degree. C. The reduced pellets in the intermediate station wherein the pellets are held for checking or temporary storage or both are at a low temperature, preferably below 100.degree. C., desirably about ambient temperature. The equipment for carrying out the method is shown schematically in the drawing. The reduction furnace 3 is watercooled by passing cooling water through cooling coil 32 on the outside. The heater winding 31 is situated outside the furnace chamber proper, which is connected via the lines 33a, b and c to a source, not shown, of an N.sub.2 /H.sub.2 gas mixture. The gas mixture leaves the furnace chamber via the line 34 and is then purified in a cleaning device 35. There, binder agents which may have been driven off are condensed and dust is separated electrostatically. The material to be sintered is loaded on transport boats, not shown, of highly heat-resistant material such as, for example, molybdenum and is placed at the inlet 1 into the transport canal 19 which goes through the whole installation. After the inlet 1, an input lock 2 is provided, which shuts the interior of the reduction furnace 3 against the outside atmosphere. After this furnace is traversed, an outlet rail 4 of similar design is provided again, which serves the same purpose. Ahead of it, this canal 19 is further provided with water cooling 12, which continues to cool the material to be sintered to room temperature after it has already cooled down in the furnace 3. After passing through the outlet lock 4, the transport boats arrive at a control station 5 which may also be designated as an intermediate storage station. There, it is ascertained, for example, that the reduction process performed in the furnace 3 has taken place properly. The intermediate storage station 5 makes possible, furthermore, different throughputs in the reduction furnace and in the following pushthrough furnace 7. The latter is again equipped with external water cooling means 72. The electric heater winding in sintering furnace 7, which makes possible sintering temperatures to maximally 1760.degree. C., is located inside the furnace chamber proper. A mixture of argon and hydrogen with controllable water vapor content is fed-in and discharged via the lines 73 and 74. The locks 6 and 8 ahead of and behind the sintering furnace 7 ensure that no harmful atmosphere gets into the interior of the transport canal 19. The water cooling 13 of the transport canal 19 takes care of cooling the finished pressed bodies which leave the furnace in sintered condition. At the outlet 9 of the transport canal 19, the transport boats can then be taken from the furnace installation and the nuclear fuel pellets can be passed on for further processing, e.g., grinding. The following examples illustrate the present invention: EXAMPLE 1 Uranium oxide/Plutonium oxide powder mixtures with 2.2 stoichiometry (oxygen-to-metal ratio) are pressed without binder to form pressed bodies in the density range of 5.5 grams per cm.sup.3. These pressed bodies are loaded into transport boats of molybdenum, each transport boat taking a pressed body weight of about 4 kg. These transport boats are then run into the reduction furnace 3 via the lock 2 as illustrated in the drawing. The furnace has a temperature profile such that the temperature increases from room temperature in the first quarter of the furnace to 1000.degree. C. This temperature is maintained over one-half the length of the furnace and then drops again to room temperature in the last quarter of the length. A total gas quantity of 35 m.sup.3 per hour of nitrogen with 8% hydrogen flows-in through the furnace via the lines 33a, b, and c. The humidity content in the entering gas is less than 10 ppm. The total gas quantity is fed into the furnace 3 in such a manner that 15 m.sup.3 per hour flow in via the line 33a at the furnace exit and 10 m.sup.3 per hour each are introduced into the hot zone by two further gas supply lines 33b and 33c. The push-through or travel velocity of the transport boats is chosen so that about 12 kg UO.sub.2 pressed bodies, i.e., 3 transport boats, get into or leave the furnace per hour. The humidity of the sinter gas leaving the furnace 3 in a collecting pipe 34 is measured continuously. If the former exceeds a value of 8000 vpm H.sub.2 O, an alarm is given and either the push-through velocity is reduced or a smaller amount of pressed bodies is loaded into the individual transport boats. After being cooled down to room temperature, the transport boats are removed from the furnace 3 and taken to the checking station or intermediate storage station 5. There, the stoichiometry is checked by sampling. If it is smaller than UO.sub.2.05, the transport boats are placed in the sintering furnace proper 7 with a temperature higher than 1600.degree. C. The push-through velocity through this furnace is controlled uniformly for all pressed bodies in such a manner that the residence times remain the same in the zone of the highest temperature and corresponds to the requirements desired for the nuclear fuel. Through this sintering furnace flows a gas mixture of argon and 8% hydrogen as well as an adjustable water content. This water content is adjusted so that the oxygen potential (hydrogen:water ratio) of the gas at the sintering temperatures is equal to the oxygen potential in the nuclear fuel pellets of the desired stoichiometry at the same temperature. The quantity of gas to be passed through is limited here to maximally 10 m.sup.3 per hour. EXAMPLE 2 Uranium oxide/Plutonium oxide powders with 2.2 stoichiometry are pressed after the addition of binder and/or lubricating agents to form pressed bodies in the density range of about 5.6 grams per cm.sup.3 and after being pressed are loaded into the transport boats. Here, too, a pressed body weight of about 4 kg is loaded per boat, and the latter are then run into the reduction furnace 3. The temperature profile of this furnace as well as the gas supply for the reduction process are the same as in Example 1. However, the push-through velocity is to be chosen in such a manner that the driving-out of the binder or lubricating agent does not lead to permanent damage at the pressed body. The upper velocity is determined simply by examining the pressed bodies at the intermediate station. The gas leaving the furnace 3 in the collecting pipe 34 is conducted through the device 35, where the binder and lubricating agent, which have been driven out and carried from the furnace by the hot gas stream, are precipitated. Likewise, dust separation of the gas stream by electrostatic means takes place there. The further treatment of the pressed bodies is the same as in Example 1. The travel-through velocities will vary, as these depend on the composition of the nuclear fuel pellets as well as on their geometrical dimensions and can be readily determined. This procedure of separating reduction processes and the sintering proper allows one to adjust and maintain optimum operating conditions for both zones of heating, so that an end product of the highest possible quality is obtained. |
claims | 1. A tool for unlatching a control rod drive shaft of a nuclear reactor vessel, comprising:a support assembly structured to receive the control rod drive shaft in a first end thereof, the support assembly having a plurality of latch fingers positioned at the first end thereof and at least one pin positioned at a second end thereof, each of the latch fingers being movable between a latched position wherein the latch finger is structured to engage and hold the control rod drive shaft when the control rod drive shaft is received in the first end and an unlatched position wherein the latch fingers are structured to not engage the control rod drive shaft when the control rod drive shaft is received in the first end; anda latching assembly, wherein the support assembly is received within the latching assembly in a manner wherein the latching assembly is moveable relative to the support assembly, the latching assembly including a first sleeve member at a first end thereof and a second sleeve member at a second end thereof, the second sleeve member having at least one inverted J-shaped slot having a first portion having a first terminal end comprising a notch, a second portion having a second terminal end, and a middle portion provided between the first portion and the second portion, the second portion being longer than the first portion, the inverted J-shaped slot being positioned such that the first terminal end is located a first distance from the latch fingers and the second terminal end is located a second distance from the latch fingers that is shorter than the first distance, wherein the at least one pin is moveably received within the at least one inverted J-shaped slot, wherein the latching assembly is movable in an unlatching manner from a latched state to an unlatched state wherein the latching assembly slides relative to the support assembly in a first direction and causes the first sleeve member to engage each latch finger and move each latch finger from the latched position to the unlatched position and wherein the latching assembly is movable in a latching manner from the unlatched state to the latched state wherein the latching assembly slides relative to the support assembly in a second direction opposite the first direction and causes the first sleeve member to engage each latch finger and move each latch finger from the unlatched position to the latched position, wherein in the latched state the at least one pin rests against the first terminal end of the inverted J-shaped slot in a manner wherein the support assembly hangs from the latching assembly, wherein during movement in the unlatching manner the latching assembly rotates relative to the support assembly in a first rotational direction and the at least one pin moves from the first terminal end of the inverted J-shaped slot to a second terminal end of the inverted J-shaped, wherein during movement in the latching manner the latching assembly rotates relative to the support assembly in a second rotational direction opposite the first rotational direction and the at least one pin moves from the second terminal end of the inverted J-shaped slot to the first terminal end of the inverted J-shaped slot during movement in the latching manner, and wherein during movement in the unlatching manner the latching assembly first slides in the second direction, then rotates in the first rotational direction and then slides in the first direction, and wherein during movement in the latching manner the latching assembly first slides in the second direction, then rotates in the second rotational direction and then slides in the first direction. 2. The tool according to claim 1, wherein the support assembly is structured to selectively grab and hold a top portion of the control rod drive shaft and disengage the control rod drive shaft from a bracket of the nuclear reactor vessel. 3. The tool according to claim 2, wherein the support assembly includes a plurality of selectively moveable button fingers structured to gab and hold the top portion of the CRDS and disengage the CRDS from a spider bracket. 4. The tool according to claim 3, wherein the button fingers are selectively pneumatically actuated. 5. The tool according to claim 1, wherein the at least one pin is at least two pins including a first pin and a second pin, wherein the at least one inverted J-shaped slot is a first inverted J-shaped slot in which the first pin is received and a second inverted J-shaped slot in which the second pin is received, wherein the first pin moves from a first terminal end of the first inverted J-shaped slot to a second terminal end of the first inverted J-shaped slot and the second pin moves from a first terminal end of the second inverted J-shaped slot to a second terminal end of the second inverted J-shaped slot during movement in the unlatching manner, and wherein the first pin moves from the second terminal end of the first inverted J-shaped slot to the second terminal end of the first inverted J-shaped slot and the second pin moves from the second terminal end of the second inverted J-shaped slot to the second terminal end of the second inverted J-shaped slot during movement in the latching manner. 6. The tool according to claim 1, wherein the first sleeve member includes a plurality of L-shaped slots extending through the first sleeve member, wherein each L-shaped slot receives a cam portion of a respective one of the latch fingers such that the cam portion extends through the first sleeve member. 7. The tool according to claim 1, further comprising a locking pin member, wherein the second sleeve member includes a latch hole, wherein the second end of the support assembly includes a receiving hole, wherein in the latched state the receiving hole is aligned with the latch hole and the receiving hole and the latch hole are structured to receive the locking pin member to prevent relative movement between the latching assembly and the support assembly. 8. The tool according to claim 7, wherein the second sleeve member includes an unlatch hole, wherein in the unlatched state the receiving hole is aligned with the unlatch hole and the receiving hole and the unlatch hole are structured to receive the locking pin member to prevent relative movement between the latching assembly and the support assembly. |
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046648801 | abstract | A debris trap is mounted within a bottom nozzle of a fuel assembly so as to capture and retain debris carried by coolant flowing from the lower core plate openings of the nuclear reactor to the fuel assembly. The trap includes a hollow enclosure disposed below the adapter plate of the bottom nozzle and between the corner legs of the nozzle. The enclosure is composed of upper and lower walls and a continuous side wall which spaces apart the upper and lower walls and interconnects their peripheries so as to form a debris capturing and retaining chamber within the hollow enclosure. The walls are composed of a material, such as wire mesh screen, which is permeable to liquid coolant but impermeable to debris. A plurality of wall sections severed from the lower wall and bent into the chamber of the enclosure define a plurality of openings into the chamber. The openings and wall sections which define them are oriented relative to the direction of coolant flow through the enclosure such that any debris which enters the chamber through the openings will be substantially detered from exiting back through the openings. Also, leaf springs are attached to opposite sides of the enclosure and engagable with a pair of corner legs of the bottom nozzle for releasably locking the trap enclosure in place in the bottom nozzle. Further, a central annular sleeve mounted between the upper and lower walls of th enclosure bolsters the structural integrity of the debris trap. |
046506349 | claims | 1. A mounting system for a television camera for use with a refueling machine having a gripper assembly, to facilitate the guidance of the gripper assembly relative to fuel assemblies of a nuclear reactor core comprising: a vertically movable mast; gripper means mounted on said movable mast; actuator means disposed within said mast for actuating said gripper means; a television camera, disposed in a housing, coaxially mounted within said mast by means of a shock absorbing system, whereby said television camera is isolated from both vertically upwardly and vertically downwardly directed shock loads; said shock absorbing system having first spring-biased means normally biased vertically downwardly, and second spring-biased means normally biased vertically upwardly, with said television camera interposed between said first and second spring-biased means; and a pair of semi-circular support rings disposed around said housing between said first and second spring-biased means, said support rings having an arcuate extent less than 180.degree. and defining diametrically-opposed, radially extending slots therebetween, said camera housing having a pair of diametrically-opposed, outwardly extending lugs thereon and each of said semi-circular support rings having a recess on the upper portion thereof, whereby said lugs can be aligned with said slots, permitting upward axial movement of said camera housing, so that said lugs can clear said support rings, and said camera housing can be rotated to dispose said lugs within said recesses and lock said camera housing in place. 2. The mounting system of claim 1 wherein each of said recesses is located at a circumferential position at about 90.degree. from said slots so that said camera housing can be rotated about 90.degree. to dispose said lugs within said recesses in the support rings. |
abstract | An analysis device for detecting fission products by measurement of a radioactivity includes a first line for carrying a liquid sample, a first detector connected to the first line and designed for measuring the radioactivity of fission products contained in the liquid sample, a second line for carrying a gas sample and a second detector connected to the second line and designed for measuring the radioactivity of fission products contained in the gas sample. The analysis device includes a separation device for separating gas from the first line carrying the liquid sample, which line has an outlet opening into the second line for gas separated from the liquid sample. The outlet opening fluidly connected to the second lines in such a manner that the gas separated from the liquid sample is suppliable as a gas sample to the second detector for measuring the radioactivity of fission products contained therein. |
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abstract | A fuel rack for supporting radioactive fuel assemblies in an underwater (or other submerged) environment that reduces the depth required for the pool to effectuate the fuel rack loading procedure. The fuel rack is specially designed to afford lateral loading. In one embodiment, the fuel rack comprises a body structure comprising at least one substantially vertically oriented elongated cell for receiving a nuclear fuel assembly, the body having a top, a bottom and a first lateral side; at least one elongated slot in the first lateral side of the body structure that forms a passageway into the cell through which a vertically oriented fuel assembly can be loaded; and means for supporting a fuel assembly within the cell in a substantially vertical orientation. In another embodiment, the invention is a method of laterally loading a fuel rack that utilizes rotation of the fuel assembly to secure the fuel assembly within its designated cell. |
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abstract | A projection optical unit images an object field in an image field. The projection optical unit includes a plurality of mirrors guides imaging light from the object field to the image field. At least two of the mirrors are arranged directly behind one another in the beam path of the imaging light for grazing incidence with an angle of incidence of the imaging light which is greater than 60°. This results in an imaging optical unit that can exhibit a well-corrected imageable field with, at the same time, a high imaging light throughput. |
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description | The present application is a national phase entry under 35 U.S.C § 371 of International Application No. PCT/EP2018/085039 filed Dec. 14, 2018, which claims priority from French Application No. 1762199 filed Dec. 15, 2017, all of which are hereby incorporated herein by reference. The present invention relates to a method for identifying the unit causing a raw water leak in a condenser consisting of several units. It finds application in industrial plants with a steam production circuit and a cooling circuit using a condenser, including thermal power plants such as nuclear power plants or fossil fuel power plants. In general, the water that feeds the cooling circuit of such a plant comes directly from a river or the sea. It is so-called raw water, as opposed to the purified water employed in other circuits. A nuclear power plant (1), for example, consists of three separate circuits (FIG. 1): The primary circuit (2) is the heat producing source, thanks to a pressurized water reactor (5-1) enclosed in a vessel (5-2). The water (6) that flows through the reactor vessel is heated to a high temperature, around 310° C. A pressurizer (7) is used to establish a pressure of about 155 bar so that the water remains in the liquid state. After passing through the steam generator (9) to release its heat, the water is recirculated by a pump (8) and heated again in the vessel (5-2). This primary circuit thus forms a closed loop. Depending on the power and the series of the nuclear power plant, its primary circuit may consist of several loops. However, it has only a single reactor (5-1) and a single pressurizer (7). The secondary circuit (3) recovers the steam (15) produced by the steam generators (9) of the loops of the primary circuit (2) to drive a set of turbines (10) to produce electricity (12) via an alternator (13). The tertiary circuit (4), also known as the cold source circuit (14), ensures the cooling of the steam (15) at the condensers (11), so as to obtain water in a liquid state, the condensate (17), which returns to the steam generator (9) by the action of the pump (18). This cooling circuit is supplied with raw water by the pump (16). On the other hand, in order to guarantee the longevity of such a plant, it is important to preserve the quality of the water circulating in the primary (2) and secondary (3) circuits, and therefore to avoid any entry of raw water and the pollutants it contains into those circuits. In particular, Na+, Ca2+ and Cl− ions are considered to be pollutants in that they are likely to cause corrosion. An entry of raw water into a steam generator (9) of the secondary circuit can be extremely disadvantageous for the plant operator. Indeed, as shown in detail in FIG. 2, it results in the intrusion of impurities (21), such as Na+, Ca2+ and/or Cl−, which, due to their low volatility, tend to concentrate in the liquid phase. This concentration phenomenon can generate localized corrosion which damages the steam generator tubes (19), within which the water (6) of the primary circuit circulates, and can eventually lead to puncture (30). Due to the pressure gradient between the primary (155 bar) and secondary (70-80 bar) circuits, such punctures lead to a leakage of potentially radioactive water into the water of the secondary circuit, located in an area of the plant that is not supposed to be exposed to radioactive risks. This incident is considered significant according to the scale established by the Nuclear Safety Authority and, as such, leads to an immediate and prolonged shutdown of the nuclear reactor, with the obligation to repair and decontaminate the secondary circuit. This results in a significant loss of production for the operator. In addition, pollution by raw water can lead to widespread corrosion in other parts of the secondary circuit, thus limiting the service life of the affected components. It is therefore imperative to monitor and detect any raw water infiltration in order to prevent the risks previously listed. The steam generator is thus equipped with a purification circuit known as SPG (supply, purge, steam generator), which has a dual function. The first role of this purge circuit, composed of filters and cationic and anionic demineralizers, is to clean the condensates before reinjecting the water into the secondary circuit. The other function of the SPG circuit is to make a direct measurement of the pollutants present in the steam generator using conductivity meters associated with cationic resins and sodium meters. It thus enables the detection of raw water entering the tertiary circuit from the condenser. A condenser of a pressurized-water power plant is subdivided into several identical units. At present, the condensers used in French nuclear power plants are typically composed of 3 to 8 units. Each unit corresponds to an elementary condenser (31) (FIG. 3) operating as a single, autonomous condenser within which heat exchange takes place between, on the one hand, the steam (15) to be condensed produced by the steam generator and, on the other, the liquid water from the cold water source, raw water (14) at atmospheric pressure. The secondary part of the condenser (32), consisting of the tubes (33) in which the steam circulates, is kept under vacuum so as to promote the condensation of the steam and thus increase the thermal efficiency of the plant. These tubes being in direct contact with the raw cooling water are susceptible to corrosion which can cause a puncture (34) and thus, due to the pressure gradient, an entry of raw water into the secondary circuit, which mixes with the condensate (17) which feeds the steam generator (9). Conventionally, each condenser unit is equipped with a water quality monitoring system (35) to detect raw water infiltration into the condensate. For this purpose, a sample of the condensate is tapped for analysis in the device consisting of an ion-exchange resin (36) and a conductivity meter (37). If, for example, raw river water has entered the condenser tubes, the condensate sample contains, in particular, Na+ and Ca2+ ions, which will be captured by the cationic resin and thus cause the departure of H+ ions which will increase the conductivity of the water. The measurement of the cationic conductivity of the water after it has passed over the ion-exchange resin through the conductivity meter is proportional to the concentration of Na+ and Ca2+ ions in the condensate. This method of pollutant detection is described in document EP0013528. The identification of the condenser unit causing a raw water leak based on the principle of cationic conductivity measurement was effective when the condensers were made of copper. Indeed, these condensers were subject to significant corrosion, causing entries of sufficient volumes of raw water to be detected by such a method. In recent years, however, more and more power plants have been equipped with stainless-steel condensers, which are less prone to corrosion. As a result, raw water leaks, when they do occur, are much less significant than with copper condensers, so that the conventional monitoring system consisting of an ion-exchange resin and a conductivity meter is no longer sensitive enough to detect them. Although small, these raw water infiltrations must nevertheless be identified in order to prevent the risks posed by corrosion. A palliative protocol has therefore been put in place, described below. The detection of an entry of raw water from the conductivity and sodium parameters measured at the steam generator via its SPG circuit remains possible. Indeed, due to the concentration phenomenon, impurities are present in sufficient amount to be detected by this monitoring system. When an anomaly is found in the measurements made in the SPG circuit, it is first necessary to validate the diagnosis of raw water infiltration from the condenser by eliminating other possible sources of pollution. Next, the operator successively isolates the plant's individual condenser units. When the unit causing the leak is isolated, the parameters measured by the SPG circuit return to their normal value. This protocol thus makes it possible to identify the faulty unit. The major disadvantage of this method is that the isolation of a condenser unit reduces the overall heat-exchange surface of the condenser and therefore requires adjusting the flow rate of the secondary circuit by lowering the reactor power, which results in production losses. Furthermore, since the SPG circuit is relatively remote from the condenser, where the raw water entry occurs, other sources of pollution from plant components located between the condenser and the steam generator may impact the SPG circuit parameters on which the diagnosis is based. Thus, these parasitic pollutions can induce an erroneous diagnosis. It is therefore important to develop an alternative procedure for determining the origin of raw water infiltration, without the need to isolate the condenser units one by one in order to avoid the drop in load inherent in such a protocol. To this end, a measurement method should be proposed which can be implemented directly at the level of the condensers, furthermore making it possible to detect leaks early and thus avoid the development of corrosion in the secondary circuit. The difficulty lies in the fact that the punctures that can occur on a stainless-steel condenser are small and therefore the corresponding raw water leakage is very small. In the nuclear power plants currently operating in France, such a leak is typically of the order of 1 to 2 L/h. The water flow rate in the condenser being roughly 700 to 800 m3/h, the leak is therefore diluted by a factor of 700 000 to 800 000. The average concentration of the preponderant ion in the cooling water used at the site of the development of the invention, Ca2+, is of the order of 50 ppm (50 mg/L). The amount to be measured is therefore theoretically about 0.07 μg/L. The most sensitive detection means currently available, such as atomic absorption spectrometry, make it possible to measure values of the order of 1 ppb (1 μg/L), which represents a value more than 14 times higher than the theoretical concentration that a leak in a stainless-steel condenser can generate. The inventors of the present invention have nevertheless succeeded in developing a method enabling a reliable and direct identification of the stainless-steel condenser unit at the origin of a low entry of raw water into a steam generator of a thermal power plant, without the need to successively isolate the condenser units from the plant. The method forming the subject matter of the present invention is based on the concentration of ions present in the raw water by trapping in a column containing an ion-exchange resin, each condenser box being provided with its own column. The nature of the ion-exchange resin depends on the origin of the raw water that feeds the condenser cooling circuit: in the case of river water, the ions capable of causing corrosion are Na+ and Ca2+ ions, which can be trapped by a suitable cationic resin; in the case of sea water, the ions capable of causing corrosion are Cl− ions, which can be trapped by a suitable anionic resin. When a certain amount of water has circulated through the column, the resin is collected and the fixed ions are released by elution with an acidic (respectively basic) solution, further allowing regeneration of the cationic (respectively anionic) resin. It is then possible to determine the concentration of ions capable of causing corrosion in the condensate that has circulated through the column of a given condenser unit by a spectroscopic method. By comparing the concentrations determined for each of the units, it is possible to identify which one is failing. This protocol can be applied as soon as a predetermined threshold or an abnormal evolution of the parameters measured in the SPG circuit of the steam generator is observed. It can be implemented with the unit on or unit off. In order to obtain, after elution of the resin, an amount of ions detectable by existing spectroscopic methods, the inventors carried out work to resize the volume of resin used compared with those used in the traditional method of measuring cationic conductivity. The volume of condensate eluted into the resin also affects the amount of ions obtained after elution. Furthermore, surprisingly, it proves that the direct use of new commercial resins did not lead to the identification of the failed condenser unit. Thus, the method forming the subject matter of the present invention requires a preliminary treatment step of the ion-exchange resin before the protocol is implemented. The present invention therefore relates to a method for identifying the unit causing a leak of raw cooling water in a condenser of a thermal power plant consisting of n units, n being an integer comprised between 2 and 15, preferably between 3 and 8, wherein each of the n units is equipped with a cartridge intended to contain an ion-exchange resin in a volume comprised between 50 and 150 mL, advantageously between 80 and 120 mL, comprising the following steps: a) for each of the n units, purifying the ion-exchange resin to be placed in the cartridge; b) for each of the n units, placing the purified ion-exchange resin obtained at the end of step a) in the cartridge; c) for each of the n units, passing a volume of condensate comprised between 500 and 1 500 L, advantageously between 800 and 1 200 L, into the cartridge containing the purified ion-exchange resin put in place in step b); d) for each of the n units, collecting the ion-exchange resin obtained at the end of step c); e) for each of the n units, regenerating the ion-exchange resin collected in step d) by elution with an aqueous regeneration solution; f) for each of the n units, collecting the eluate obtained at the end of step e) followed by determining the nature of the ionic species present in said eluate and the amount of each ionic species present in said eluate; and g) for each of the ionic species identified in step f), comparing the amounts of said ionic species determined in each of the n eluates. For the purposes of the present invention, “raw cooling water” means the cold source that feeds the condenser, which consists of unpurified river or sea water. A “condenser unit” means, for the purposes of the present invention, an elementary condenser, it being understood that the condenser of a thermal power plant is compartmentalized into several identical elementary condensers. For the purposes of the present invention, “condensate” means water in the liquid state resulting from the condensation of water vapor from the steam generator. Said condensate may also contain raw cooling water as a result of infiltration following a puncture in a condenser tube, said puncture may in particular result from a corrosion phenomenon. Pollutants are then present in the condensate which are likely to cause corrosion. An “ion-exchange resin” is a solid material generally in the form of beads composed of a “polymer matrix” on which are grafted positively or negatively charged functional groups that will allow an “ion exchange”. The “polymer matrix” according to the invention may be gel-type or macroporous, advantageously gel-type. It may in particular be a matrix of polystyrene, polystyrene-divinylbenzene copolymer or cross-linked polyacrylate. The average diameter of the resin beads according to the invention is comprised between 0.2 mm and 1.2 mm, notably between 0.3 mm and 0.8 mm, in particular between 0.4 mm and 0.7 mm. The resin beads according to the invention are further characterized by a uniformity coefficient less than or equal to 1.8, in particular less than or equal to 1.5, preferably less than or equal to 1.2. “Ion exchange” is a process in which ions of a certain charge contained in a solution are removed from that solution by adsorption onto a solid material, the ion exchanger, to be replaced by an equivalent amount of other ions of the same charge emitted by the solid. A distinction is made between cationic and anionic resins. A “cationic resin” has negatively charged functional groups. When an aqueous solution containing ions is circulated over such a resin, the cations initially present as counterions to the functional groups in order to ensure the electroneutrality of the resin are gradually replaced by the cations present in the eluent solution until the resin is saturated. The progressive saturation of a cationic resin initially containing H+ ions by exchange with Na+ ions is shown in FIG. 4. An “anionic resin” has positively charged functional groups, with the understanding that it may be a partial charge. When an aqueous solution containing ions is circulated over such a resin, the anions initially contained in the resin will gradually be replaced by the anions present in the eluent solution, until the resin is saturated. The “exchange capacity” of an ion-exchange resin, expressed in equivalents per liter, is the capacity to retain ions until the resin is saturated. Its “total exchange capacity” corresponds to the equivalent concentration of functional groups, and thus the number of monovalent exchangeable ions per unit volume. An ion-exchange resin according to the invention has a total exchange capacity greater than 1.0 eq/L, preferably greater than 1.5 eq/L, advantageously greater than 2.0 eq/L. A saturated ion-exchange resin can be regenerated by elution using a “regeneration solution” containing ions of the same charge as those adsorbed on the resin. The “regeneration” of a cationic resin can be carried out using an aqueous regeneration solution consisting of a solution of a mineral acid, preferably a strong acid (pKa≤0), such as hydrochloric acid, nitric acid or sulfuric acid. The gradual regeneration of a cationic resin saturated with Na+ ions is shown in FIG. 5. The regeneration of an anionic resin can be carried out using an aqueous regeneration solution consisting of an aqueous solution of a base, preferably a strong base (pKa≥14), such as soda. An ion-exchange resin can be selective for certain ions, i.e. preferentially fix certain ions. For the purposes of the present invention, “purification of the ion-exchange resin” means an elution process to remove impurities from the resin. Case of Raw River Water When the cooling circuit is fed with river water, the pollutants capable of causing corrosion include Na+ and Ca2+ ions. The “ion-exchange resin” used in the method according to the invention then consists of a cationic resin. The resin is then in the form of polymer beads, in particular polystyrene or polystyrene-divinylbenzene copolymer, on which negatively charged cation-exchange functional groups are grafted. The resin preferably has a high affinity for Na+ and Ca2+ ions. Advantageously, the resin is selective for Na+ and Ca2+ ions. It can be a strongly acidic cationic resin with, for example, SO3− groups as exchange functional groups or a weakly acidic cationic resin with, for example, CO2− groups as exchange functional groups. Advantageously, it is a strongly acidic resin. In a particular embodiment, the ion-exchange resin is a strongly acidic cationic resin, which consists of polystyrene or polystyrene-divinylbenzene copolymer beads on which are grafted SO3− groups and having a uniformity coefficient less than or equal to 1,8, in particular less than or equal to 1.5, preferably less than or equal to 1.2, the total exchange capacity of the resin being greater than 1.0 eq/L, preferably greater than 1.5 eq/L, advantageously greater than 2.0 eq/L. Advantageously, the counterions of the SO3− groups initially present in the resin before the implementation of the method forming the subject matter of the present invention are H+ ions. In particular, prior to the implementation of the method forming the subject matter of the present invention, at least 99% of the SO3− groups have an H+ ion as counterion. Advantageously, prior to the implementation of the method forming the subject matter of the present invention, the concentration of Na+ ions in the resin is less than 100 ppm, preferably less than 70 ppm, preferentially less than 50 ppm. Advantageously, prior to the implementation of the method forming the subject matter of the present invention, the concentration of Ca2+ ions in the resin is less than 100 ppm, preferably less than 70 ppm, preferentially less than 50 ppm. Advantageously, prior to the implementation of the method forming the subject matter of the present invention, the concentration of each of the cations present in the resin is less than 100 ppm, preferably less than 70 ppm, preferentially less than 50 ppm. The cationic resin may in particular be a resin specially intended for nuclear use. Thus, the cationic resin can advantageously be selected from the resins Amberlite IRN97 H, IRN77 and IRN99 marketed by Dow and the resins Purolite NRW1000, NRW1100, NRW1160, NRW1180 and NRW160 marketed by Purolite. In particular, the cationic resin can be the resin Amberlite IRN97 H. As mentioned above, it is necessary to proceed during step a) of the method according to the invention to the purification of the cationic resin intended to be placed in the cartridge which equips each condenser unit. Purification is carried out by elution using an acidic solution, preferably a strong acid, such as hydrochloric acid, nitric acid or sulfuric acid. The mass percentage of the acid in solution is comprised between 1 and 50%, notably between 5 and 30%, in particular between 10 and 20% by weight relative to the total weight of the solution. The elution operation consists in pouring onto the resin a volume of acidic solution at least 2 times as large, in particular at least 4 times as large, advantageously at least 5 times as large as the volume of resin. Advantageously, the acidic purification solution has an Na+ ion concentration of less than 1 ppb, notably less than 0.5 ppb, in particular less than 0.2 ppb. Advantageously, the acidic purification solution has a Ca2+ ion concentration of less than 1 ppb, notably less than 0.5 ppb, in particular less than 0.2 ppb. Advantageously, each cation present in the acidic purification solution has a concentration of less than 1 ppb, notably less than 0.5 ppb, in particular less than 0.2 ppb. In a preferred embodiment, after the purification step a), the concentrations of Na+ and Ca2+ ions in the resin are less than 1 ppb, notably less than 0.5 ppb, in particular less than 0.2 ppb. In step c) of the method according to the invention, the H+ ions present in the purified cationic resin obtained at the end of step a) and placed in step b) are exchanged with the Na+ and/or Ca2+ ions of the condensate until the resin is saturated. Step e) of regenerating the cationic resin can be carried out by elution using an aqueous regeneration solution consisting of a solution of a mineral acid, preferably a strong acid such as hydrochloric acid, nitric acid or sulfuric acid. The mass percentage of the acid in the regeneration solution is comprised between 1 and 50%, notably between 5 and 30%, in particular between 10 and 20% by weight relative to the total weight of the solution. In particular, elution consists in pouring onto the resin a volume of acidic solution at least 2 times as large, in particular at least 4 times as large, advantageously at least 5 times as large as the volume of resin. Advantageously, the regeneration solution has an Na+ ion concentration of less than 1 ppb, notably less than 0.5 ppb, in particular less than 0.2 ppb. Advantageously, the regeneration solution has a Ca2+ ion concentration of less than 1 ppb, notably less than 0.5 ppb, in particular less than 0.2 ppb. Advantageously, each cation present in the regeneration solution has a concentration of less than 1 ppb, notably less than 0.5 ppb, in particular less than 0.2 ppb. In step f) of the method according to the invention, the amounts of Na+ and/or Ca2+ ions present in the eluate obtained at the end of step e) are determined. This measurement can be carried out by spectroscopic methods well known to the skilled person, such as atomic absorption spectroscopy. Case of Raw Sea Water When the cooling circuit is fed with river water, the pollutants capable of causing corrosion include Cl− ions. The “ion-exchange resin” used in the method according to the invention then consists of an anionic resin. The resin is then in the form of polymer beads, in particular polystyrene or polystyrene-divinylbenzene copolymer beads, on which anion-exchange functional groups are grafted. Preferably, the resin has a high affinity for Cl− ions. Advantageously, the resin is selective for Cl− ions. It may be a strongly basic anionic resin having N(RR′R″)+ ammonium exchange functional groups, wherein R, R′ and R″ are identical or different (C1-C6)alkyl groups. It may be a weakly basic anionic resin having as exchanger functional groups amine groups NRR', wherein R and R′ are identical or different (C1-C6)alkyl groups. For the purposes of the present invention, a “(C1-C6)alkyl” group is a saturated, linear or branched hydrocarbon chain containing 1 to 6, preferably 1 to 4, carbon atoms. By way of example, mention may be made of methyl and ethyl groups. In a particular embodiment, the ion-exchange resin is a strongly basic anionic resin, which consists of polystyrene or polystyrene-divinylbenzene copolymer beads on which ammonium groups are grafted, such as N(CH3)3+ groups and having a uniformity coefficient less than or equal to 1.8, in particular less than or equal to 1.5, preferably less than or equal to 1.2, the total exchange capacity of the resin being greater than 1.0 eq/L, preferably greater than 1.5 eq/L, advantageously greater than 2.0 eq/L. Advantageously, the counterions of the ammonium groups initially present in the resin before the implementation of the method forming the subject matter of the present invention are HO− ions. In particular, prior to the implementation of the method forming the subject matter of the present invention, at least 95% of the ammonium groups have a HO− ion as counterion. Advantageously, prior to the implementation of the method forming the subject matter of the present invention, the concentration of Cl− ions in the resin is less than 100 ppm, preferably less than 70 ppm, preferentially less than 50 ppm. Advantageously, prior to the implementation of the method forming the subject matter of the present invention, the concentration of each of the anions present in the resin is less than 100 ppm, preferably less than 70 ppm, preferentially less than 50 ppm. The anionic resin may in particular be a resin specially intended for nuclear use. Thus, the anionic resin can advantageously be chosen from the resin Amberlite IRN78 marketed by Dow and the resins Purolite NRW4000, NRW6000, NRW7000, NRW8000, NRW5010, NRW5050 and NRW5070 marketed by Purolite. As mentioned above, it is necessary to proceed in step a) of the method according to the invention to the purification of the anionic resin intended to be placed in the cartridge which equips each condenser unit. Purification is carried out by elution using a basic solution, preferably a strong basic solution, such as soda. The mass percentage of the base in solution is comprised between 1 and 50%, notably between 5 and 30%, in particular between 10 and 20% by weight relative to the total weight of the solution. The elution operation consists in pouring onto the resin a volume of basic solution at least 2 times as large, in particular at least 4 times as large, advantageously at least 5 times as large as the volume of resin. Advantageously, the basic purification solution has a Cl− ion concentration of less than 1 ppb, notably less than 0.5 ppb, in particular less than 0.2 ppb. Advantageously, each anion present in the basic purification solution has a concentration of less than 1 ppb, notably less than 0.5 ppb, in particular less than 0.2 ppb. In step c) of the method according to the invention, the HO− ions present in the purified anionic resin obtained at the end of step a) and placed in step b) are exchanged with the Cl− ions of the condensate until the resin is saturated. Step e) of regeneration of the anionic resin can be carried out by elution using an aqueous regeneration solution consisting of a basic solution, preferably a strong base such as soda. The mass percentage of the base in the regeneration solution is comprised between 1 and 50%, notably between 5 and 30%, in particular between 10 and 20% by weight relative to the total weight of the solution. In particular, elution consists in pouring onto the resin a volume of basic solution at least 2 times as large, in particular at least 4 times as large, advantageously at least 5 times as large as the volume of resin. Advantageously, the regeneration solution has a Cl− ion concentration of less than 1 ppb, notably less than 0.5 ppb, in particular less than 0.2 ppb. Advantageously, the regeneration solution has a Cl− ion concentration of less than 1 ppb, notably less than 0.5 ppb, in particular less than 0.2 ppb. Advantageously, each anion present in the regeneration solution has a concentration of less than 1 ppb, notably less than 0.5 ppb, in particular less than 0.2 ppb. In step f) of the method according to the invention, the amounts of Cl− ions present in the eluate obtained at the end of step e) are determined. This measurement can be carried out by the spectroscopic methods well known to the skilled person. The present invention is illustrated by the following non-limiting figures and examples. Implementation of the Method According to the Invention: The protocol described below was implemented on a nuclear power plant condenser consisting of 7 units, the cooling circuit of which is fed by raw river water containing Na+ and Ca2+ ions. The different steps are shown in FIG. 6. The cationic resin used is amberlite IRN97 H. A resin purification operation is carried out by pouring 500 mL of a hydrochloric acidic solution over 100 mL of resin. The solution used is a 15% dilution of Suprapur® hydrochloric acid marketed by Merck, the dilution being carried out with demineralized water having an Na+ and Ca2+ concentration of less than 1 ppb. A cartridge containing 100 mL of previously purified cationic resin is placed on each condenser unit. Water from the secondary circuit (a) is passed over the cartridge containing 100 mL of purified cationic resin, the cations contained in the water are retained on the resin (b). Once sufficient condensate has passed over the resin (about 1 m3), the resin is recovered and transferred (c) to a laboratory glass column (d). The resin in the glass column is then eluted (acid is circulated over it) by a concentrated acidic solution containing H+ ions (e). The H+ ions will replace the fixed cations (g and h) on the resin. The cations thus removed (f) will be recovered and measured by a specific apparatus (i). The elution operation consists of pouring 500 mL or more of acid over 100 mL of resin. The acidic solution used is the same as that used for the purification operation. The acid flows through the resin at a rate of one to two drops per second. The eluate is recovered from the resin in 100 mL fractions. Calcium is dosed onto each fraction collected by atomic absorption spectrometry. These cations come from the known volume of condensate with which the resin was eluted. It is therefore possible to determine the concentration of Ca2+ ions present in the condensate for each unit. The results obtained are shown in FIG. 7a. Comparative Example The same protocol is reproduced, without purifying the cationic resin. The results obtained are shown in FIG. 7b. |
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description | This application claims the benefit of U.S. Provisional Applications Ser. Nos. 61/545,253 and 61/545,262, both filed Oct. 10, 2011, which provisional applications are incorporated herein by reference. This disclosure relates to scintillator materials used for detecting ionizing radiation, such as X-rays, gamma rays and thermal neutron radiation, in security, medical imaging, particle physics and other applications. This disclosure relates particularly to rare-earth metal halide scintillator materials. Certain arrangements also relate to specific compositions of such scintillator material, method of making the same and devices with such scintillator materials as components. Scintillator materials, which emit light pulses in response to impinging radiation, such as X-rays, gamma rays and thermal neutron radiation, are used in detectors that have a wide range of applications in medical imaging, particle physics, geological exploration, security and other related areas. Considerations in selecting scintillator materials typically include, but are not limited to, luminosity, decay time, emission wavelengths, and stability of the scintillation material in the intended environment. While a variety of scintillator materials have been made, there is a continuous need for superior scintillator materials. Metal Halides, especially rare earth metal halides such as LaBr3, LaCl3, CeBr3, CeCl3 and LuI3, are scintillator compositions known from their good energy resolution and relatively high light output. The main disadvantage of these materials is their extremely high solubility in water. Hygroscopicity is one of the main reasons that slows down the process of commercialization of these compounds. Crystal growth processes, following a multistage purification, zone refining and drying all require very well controlled atmosphere with depleted content of water and oxygen. Moreover, handling and post-growth processing of these materials has to be performed in an ultra-dry environment to avoid degradation of materials. Furthermore, many of these compounds are light sensitive and thus require additional handling steps. In addition, such materials often can be used only in the hermetic package that prevents them from degradation due to the hydration effects. Therefore, it is desirable to improve or develop new scintillator materials with significantly lower solubility in water (i.e., lower hygroscopicity). This disclosure relates generally to rare-earth metal halide scintillator materials and method of making such scintillator materials. In one arrangement, the rare-earth metal halide scintillator materials have compositions with reduced hygroscopicity. Compositions in specific implementations include three group of elements: Lanthanides, (La, Ce, Lu, Gd or V), elements in group 17 of the periodic table of elements (CI, Br and I) and elements of group 13 (B, AI, Ga, In, TI), and any combination of these elements. A further aspect of the present disclosure relates to a method of making chloride scintillator materials of the above-mentioned compositions. In one example, high-purity starting halides (such as TlBr and CeBr3) are mixed and melted to synthesize a compound of the desired composition of the scintillator material. A single crystal of the scintillator material is then grown from the synthesized compound by the Bridgman method (or Vertical Gradient Freeze (VGF) method), in which a sealed ampoule containing the synthesized compound is transported from a hot zone to a cold zone through a controlled temperature gradient at a controlled speed to form a single-crystalline scintillator from molten synthesized compound. Another aspect of the present disclosure relates to a method of using a detector comprising one of the scintillation materials described above for imaging. This disclosure is related to the new compositions of rare earth metal halides where the change in the character of the compounds is achieved by adding of elements from group 13 of Periodic Table of Elements. These elements may create covalent bonds with metal halides that result in their lower hygroscopicity. A good example of group-13 compounds is TlBr, which is known for being insoluble in water. Introduction of Tl into the rare earth metal halides, such as LaBr3 and CeBr3, results in creation of TI-Br covalent bonds. These bonds change the character of these compounds from being “Hard Acid-Hard Base” to “Soft Acid-Soft Base.” The physical forms of the scintillator substance include, but are not limited to, crystal, polycrystalline, ceramic, powder or any of composite forms of the material. A reduction in the hygroscopicity is achieved by co-doping and/or changes in the stoichiometry of a scintillator substance. These changes may be achieved by stoichiometric admixture and/or solid solution of compounds containing elements from group-13 periodic table. One way of the implementation of this innovation is a codoping with one or more group-13 elements in concentrations that do not alter significantly the symmetry of the crystal lattice of the scintillator of choice. Another way includes a complete modification of the crystal structure of the scintillator composition by stoichiometric change or solid solution of scintillator compounds and other compounds containing at least one of group-13 elements. In these cases, new scintillator materials are created with significantly reduced hygroscopicity. The present disclosure includes, but is not being limited to, the following families of metal halides compositions described by general chemical formulas: A′(1−x)B′xCa(1−y)EuyC′3 (1), A′3(1−x)B′3xM′Br6(1−y)Cl6y (2), A′(1−x)B′xM′2Br7(1−y)C′7y (3), A′(1−x)B′xM″(1−y)EuyI3 (4), A′3(1−x)B′3xM″(1−y)EuyI5 (5), A′(1−x)B′xM″2(1−y)Eu2yI5 (6), A′3(1−x)B′3xM′2Cl7 (7), A′(1−x)B′xM′2Cl7 (8), and M′(1−x)B′xC′3 (9), wherein: A′=Li, Na, K, Rb, Cs or any combination thereof, B′=B, Al, Ga, In, Tl or any combination thereof, C′=Cl, Br, I or any combination thereof, M′ consist of Ce, Sc, Y, La, Lu, Gd, Pr, Tb, Yb, Nd or any combination thereof, M″ consists of Sr, Ca, Ba or any combination of thereof, 0≤x≤1, and 0≤y≤1. In a particular, non-limiting, example, thallium (Tl) is introduced into the crystallographic lattice of LaBr3 compound (formula 9). In this specific example, a strong Tl—Br covalent bond (as opposed to ionic bond in LaBr3) is created that significantly reduces the reactivity of the compound with water. In the higher concentration of Tl it is possible to create scintillator materials with altered crystallographic lattice. That includes also a stoichiometry change in the crystal itself. The strength of Tl—Br bond is demonstrated in TlBr compound that is known from significantly lower hygroscopicity in comparison to the other rare-earth metal halides. The expected changes in solubility can be explained based on the HSAB concept, explained in more detail below. Moreover, introduction of the elements from group-13 into the crystal structure of rare-earth metal halides often improves scintillation characteristics of these materials. Addition of Tl as a codopant or stoichiometric admixture to certain compositions of rare-earth metal halides creates very efficient scintillation centers. These centers contribute to the scintillation light output. In addition, using compounds of group-13 elements can favorably increase the density of the material. Improvement in the density is particularly important in radiation detection applications. The new scintillator materials have applications in Positron Emission Tomography (PET), Single Photon Emission Computed Tomography (SPECT), Computerized Tomography (CT), and other applications used in homeland security and well logging industry. This disclosure also relates to the method of growing scintillator that includes crystallization of the melted or dissolved scintillator compounds under controlled environment. The changes in solubility of new rare-earth metal halides scintillators disclosed herein may be understood based on HSAB concept. The HSAB is an acronym for “Hard and Soft Acids and Bases” known also, as the Pearson acid-base concept. This concept attempts to unify inorganic and organic reaction chemistry and can be used to explain in qualitative rather than quantitative way the stability of compounds, reaction mechanisms and pathways. The concept assigns the terms ‘hard’ or ‘soft’, and ‘acid’ or ‘base’ to variety of chemical species. ‘Hard’ applies to species which are small based on their Ionic radii, have high charge states (the charge criterion applies mainly to acids, to a lesser extent to bases), and are weakly polarizable. ‘Soft’ applies to species which are big, have low charge states and are strongly polarizable. Polarizable species can form covalent bonds, whereas non-polarizable form ionic bonds. See, for example, (1) Jolly, W. L., Modern Inorganic Chemistry, New York: McGraw-Hill (1984); and (2) E.-C. Koch, Acid-Base Interactions in Energetic Materials: I. The Hard and Soft Acids and Bases (HSAB) Principle-Insights to Reactivity and Sensitivity of Energetic Materials, Prop., Expl., Pyrotech. 30 2005, 5. Both of the references are incorporated herein by reference. In the context of this disclosure the HSAB theory helps in understanding the predominant factors which drive chemical properties and reactions. In this case, the qualitative factor is solubility in water. On the one hand, water is a hard acid and hard base combination, so it is compatible with hard acid and bases. Thallium bromide is, on another hand, a soft acid and soft base combination, so it is not soluble in water. According to the HSAB theory, soft acids react faster and form stronger bonds with soft bases, whereas hard acids react faster and form stronger bonds with hard bases, all other factors being equal. Hard acids and hard bases tend to have the following characteristics: small atomic/ionic radius high oxidation state low polarlzabllity high electronegativity (bases) Examples of hard acids include: H+, light alkali ions (for example, Li through K all have small ionic radius), Ti4+, Cr3+, Cr6+, BF3. Examples of hard bases are: OH−, F−, Cl−, NH3, CH3COO− and CO32−. The affinity of hard acids and hard bases for each other is mainly ionic in nature. Soft acids and soft bases tend to have the following characteristics: large atomic/ionic radius low or zero oxidation state high polarizability low electronegativity Examples of soft acids are: CH3Hg+, Pt2+, Ag+, Au+, Hg2+, Hg22+, Cd2+, BH3 and group-13 in +1 oxidation state. Examples of soft bases include: H−, R3P, SCN− and I−. The affinity of soft acids and bases for each other is mainly covalent in nature. There are also borderline cases identified as borderline acids for example: trimethylborane, sulfur dioxide and ferrous Fe2+, cobalt Co2+, cesium Cs+ and lead Pb2+ cations, and borderline bases such as bromine, nitrate and sulfate anions. Generally speaking, acids and bases interact and the most stable interactions are hard-hard (ionogenic character) and soft-soft (covalent character). In the specific case presented as an example compounds such as LaBr3 and TlBr have the following elements to consider following reaction with water: La+3, Br−, Tl+, H−, OH−. La+3: This is a strong acid. High positive charge (+3) small ionic radius. Br−: This is a soft base. Large ionic radius small charge (−1). Tl+: This is a soft acid. Low charge and large ionic radius. H+: This is a hard acid. Low ionic radius and high charge density. OH−: This is a hard base. Low charge, small ionic radius. Thus the reaction of LaBr3 and water takes place in according to the following scheme:[La+3,Br−]+[H+, OH−]→[La+3, OH−]+[H+, Br]. The left hand side of the equation has two components that are being mixed. The right hand side represents products after mixing. One can see that the strong acid La+3 with the strong base OH−, are joined together because it makes a strong acid and base combination. The Br−is driven from the La+3 and thus it is complexed with H+, forming hydrobromic acid. The reaction of TlBr with water following the scheme:[Tl+,Br−]+[H+, OH−]→[Tl+, Br−]+[H+, OH−]. In this case, Tl+ and Br− are favored because they are a combination of soft-soft acid and base. While the H+ and OH− are hard acid and base combination. The TlBr is a covalent compound and will dissolve in covalent solvents. Therefore, in the case of LaBr3, the hard acid La+3 “seeks” out OH−, resulting in a high reactivity in water. In contrast, TlBr (soft-soft) does not “seek” water (and vice versa). The result is a low degree of interaction, including solubility with water. In the examples given above in this disclosure, the addition of TlBr as a co-dopant or in stoichiometric amounts reduces the hygroscopicity of the LaBr3. A further aspect of the present disclosure relates to a method of making scintillator materials of the above-mentioned compositions. In one example, high-purity starting compounds (such as LaBr3 and TlBr) are mixed and melted to synthesize a compound of the desired composition of the scintillator material. A single crystal of the scintillator material is then grown from the synthesized compound by the Bridgman method (or Vertical Gradient Freeze (VGF) method), in which a sealed ampoule containing the synthesized compound is transported from a hot zone to a cold zone through a controlled temperature gradient at a controlled speed to form a single-crystalline scintillator from molten synthesized compound. Thus, rare-earth metal halide scintillation materials with improved moisture resistance, density and/or light output can be made with the addition of group-13 elements such as Tl. Because many embodiments of the invention can be made without departing from the spirit and scope of the invention, the invention resides in the claims hereinafter appended. |
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047602686 | claims | 1. A container for medium or low activity radioactive waste comprising a drum which has an axis and a side wall concentric to said axis and which is open at an upper axial end thereof, and a cover for closing said open end of the drum, said container being characterized by: A. said cover comprising an upper and a lower disc-like horizontal plate and a plurality of plate-like baffle members between said plates that connect them in spaced apart and parallel relationship, said lower plate and baffle members being receivable within the drum in a closed condition of the cover wherein said upper plate overlies the upper end of the drum all around the same, B. said drum having a circumferential ledge near its upper end that defines a downwardly facing shoulder which projects radially inwardly from said side wall; C. means carried by said baffle members at their radially outer ends providing a plurality of substantially resilient tongues at circumferentially spaced intervals around the cover, each said tongue projecting obliquely radially outwardly and axially upwardly to have a free end portion which is deflected radially inwardly by said ledge as the cover is moved down to its closed condition and which lockingly engages under said shoulder when the cover reaches said condition; and D. said cover further comprising a central coaxial ferrule extending between said plates, said ferrule a cup-shaped receptacle coaxially secured to said ferrule and surrounding the same near said upper end thereof, said receptacle having 2. The container of claim 1, further characterized by: said upper plate having an elastically deformable peripheral portion capable of flexing to bear against the upper end of said drum under resilient bias all around the same. 3. The container of claim 2, further characterized by: each said baffle member having an upper edge portion which is spaced below said peripheral portion of the upper plate to leave the same free for flatwise flexing. 4. The container of claim 3, further characterized in that: each said baffle member has a notch in its upper edge, intermediate the radially inner and radially outer ends of the baffle member, to provide for communication between said sectors at a level downwardly adjacent to the upper plate. 5. The container of claim 1, further characterized by: a gasket secured to the underside of said upper plate, around the peripheral portion thereof, for sealingly engaging the upper end of the drum. 6. The container of claim 1, further characterized by: |
046817271 | summary | BACKGROUND OF THE INVENTION The invention relates to processes for the commercially practical production of radiopharmaceutical activities and, more particularly, relates to a process for the production of astatine-211 (At-211). For about the last forty years, there has been an interest in exploring the potential of At-211 for therapeutic biomedical applications. At-211 decay results in nearly pure alpha particle emissions. The radiotoxicity associated with such alpha particle emissions and the associated emissions resulting from the decay of its daughter, polonium-211 (Po-211), which has a half-life of only about 0.56 seconds, led to early recognition of the potential for such beneficial therapeutic applications. At-211 has a half-life of 7.21 hours, which is sufficiently long to enable its practical production, chemical synthesis, transportation, quality control and appropriate biological application in a number of useful radiopharmaceutical treatments of certain diseases. Heretofore, research and development work relating to such radiopharmaceutical treatment procedures has largely focused on the preparation and in vivo evaluation of labelled antibodies, proteins, drugs and inorganic colloids. Such research has usually concentrated on the production of At-211 as a source of Astatine for studies of the physical and chemical properties of the element. In addition to the interest generated by the established and potential uses of At-211 in therapeutic radiopharmaceuticals, increasing attention is being received by At-211 due to its potential use in radiation synovectomy and for fundamental studies in cell biology. It has been demonstrated, for example, that At-211 Tellurium-colloid can be curative, without undue toxicity, in mice bearing Ascites tumor cells. Among currently available alpha particle emitters, At-211 is particularly promising for radiopharmaceutical therapeutic uses, because it decays by a double branched pathway into lead (Pb-207), directly by alpha emission, and indirectly through electron capture into polonium (Po-211), which in turn decays almost spontaneously by alpha emission. In the past, radiochemical separation of At-211 has been found to be generally unreliable due to the low and variable recovery yields attainable with such processes. If an At-211 radiopharmaceutical is to be developed for clinical application, it will be necessary to develop a production process in which small controlled volumes of At-211, in specified chemical forms, can be produced more reliably and consistently. Preferably, such a process would be readily controllable to produce desirable small volumes of selected activities in a solution of solvents that is compatible with preselected radiopharmaceutical procedures in which the activities are to be used. More specifically, it would be desirable to have such a process in which a choice of solvents can be effectively used in the production of a desired At-211 radiopharmaceutical. OBJECTS OF THE INVENTION A major object of the invention is to provide a process for reliably and consistently producing an At-211 radiopharmaceutical in a desirably small controllable volume. Another object of the invention is to produce an At-211 radiopharmaceutical by a process that includes the option of eluting At-211 with a choice of solvents, any of which are compatible with subsequently desired radiopharmaceutical procedures in which the At-211 radiopharmaceutical is to be utilized. A further object of the invention is to provide a process for reliably producing an At-211 radiopharmaceutical in the chemical form as Astatide and in a desirably small controlled volume. Still another object of the invention is to provide a process that utilizes a novel one-step distillation and collection procedure for producing At-211. Yet another object of the invention is to provide a process for isolation of At-211 by distilling it from a heated bismuth target that has been irradiated with alpha particles, collecting the vapor condensate in a silica gel, and eluting At-211 from the silica gel with a controllable volume of eluent that contains a solvent, which is preselected to be compatible with a given radiopharmaceutical procedure in which the At-211 is to be used. A still further object of the invention is to provide a process for isolating At-211 from a Bismuth target without requiring the use of other chemicals, which frequently are a source of contaminants, thereby to avoid the introduction of contaminants into the isolated At-211. Another object of the invention is to provide a process for isolating At-211, wherein the vaporization of bismuth from a target is suppressed by using a selected choice of dry carrier gases, which is effective to prevent Bi metal from contaminating the At-211 isolated by the process. Additional objects and advantages of the invention will become apparent to those skilled in the art from the description of it presented herein, considered in conjunction with the accompanying drawings. SUMMARY OF THE INVENTION In one preferred arrangement of the invention a At-211 radiopharmaceutical is produced by a process in which a target of irradiated bismuth is heated within a critical range of temperatures, for a predetermined period of time, in a suitable still, while passing a dry carrier gas through the still to carry At-211 vapor evolved from the Bi target to a condenser which has a condensate collector that is effective to remove essentially all of the At-211 from the carrier gas. Subsequently, as a safety precaution, the carrier gas is passed through a series of filters to remove remaining traces of the At-211 from it. A small controlled volume of eluent, which is preselected from a choice of solvents, is used to elute At-211 from the condensate collector. Finally, the controlled volume of eluent containing the At-211 is eluted from the condensate collector and held for use in a selected radiopharmaceutical procedure. |
060410914 | summary | BACKGROUND OF THE INVENTION The present invention relates to a control rod for a nuclear reactor, and more particularly, to a control rod for a nuclear reactor of the long-life type having an improved mechanical soundness in a boiling water reactor. A control rod for a boiling water reactor (BWR) has usually four wings formed by housing neutron absorber plates in a plurality of long sheaths having a deep U-shaped cross-section. A leading end structural member is provided at an insertion leading end portion of each of the wings, or a terminal end structural member, at an insertion terminal end portion thereof, and the U-shaped openings of the sheaths in each of four wings are secured to an integral type center structural member (known also as a tie rod) having a cruciform cross-section to provide a cruciform sectional configuration or arrangement. In a conventional control rod, the sheath is made of stainless steel (S.S such as SUS, hereinafter abbreviated as "SUS"), and a SUS tube having a diameter of 5 mm filled with boron carbide (B.sub.4 C) powder has been employed as a neutron absorbing rod. Boron (B) has however a short nuclear life because it reacts with neutron to generate helium (He) and lithium (Li), resulting in a degraded neutron absorbing ability, and helium causes an increase in internal pressure, leading to a decrease in soundness of the SUS tube and hence to a shorter mechanical and physical service life. In order to provide a control rod having a long service life, there has been used a long-life type control rod manufactured by replacing a conventional neutron absorbing rod partially or totally by hafnium (hereinafter abbreviated as "Hf") which is a long-life type neutron absorber. Since Hf has a large specific gravity (density) as about 13, an Hf rod having the same cross-section as a conventional neutron absorbing rod using boron carbide results in a weight about 1.5 times as large as the control rod as a whole, although the neutron absorbing ability (reactivity value) is substantially the same, making it impossible to back-fit it into a nuclear reactor in operation. As a counter-measure, Japanese Patent Laid-open (KOKAI) Publication No. HEI 1-34358 "Control Rod for Nuclear Reactor" proposes an Hf control rod of the type known as a trap type in which Hf is formed into a plate shape, and two Hf plates are arranged opposite to each other with a gap for introduction of water. Further, in view of the fact that in about the terminal end side half of a control rod for the BWR, upon insertion thereof into the reactor core, a decreased neutron absorbing ability would cause no inconvenience in the control of the BWR, a control rod configured so as to use a smaller Hf content in the portion on the insertion terminal end side than in the portion on the insertion leading end side is proposed in Japanese Patent Laid-open Publication No. HEI 7-3468 "Control Rod for Nuclear Reactor". Regarding the long-life control rod having the trap structure using the Hf plate, excellent results have already been achieved in many BWRs, and it is the usual practice to set a short service life for maintenance purposes. When setting a longer service life, it becomes now clearer than ever that it is effective to improve mechanical strength of SUS structural members such as a sheath in the control rod. FIGS. 19 to 21 illustrate an outline of an Hf trap type control rod, in which FIG. 19A is a partially cutaway perspective view, FIG. 19B is a sectional view of a wing and FIG. 19C is a perspective view of a load supporting member (also referred to as a "load supporting spacer" of "top spacer"). FIG. 20A is a partially cutaway front view of a sheath shown in FIG. 19A, and FIG. 20B illustrates an example of thickness of an Hf plate which is a neutron absorber plate of a neutron absorbing material attached in the interior of a sheath, as illustrated in a distribution diagram in the control rod insertion/withdrawal direction which is the sheath longitudinal direction. FIG. 21A is a partially enlarged front view of FIG. 20A, FIG. 21B is an enlarged front view of a pair of Hf plates shown in FIG. 21A, and FIG. 21C is a sectional view of FIG. 21C taken along the line XXIC--XXIC of FIG. 21B. Referring to these figures, a long-life type control rod 1 has a cruciform section with four wings 2, and a leading end structural member 4 integral with a handle 3 is secured to the insertion leading end portion into the reactor core, and a terminal end structural member 5 is secured to an insertion terminal end portion. Further, a cruciform integral type center structural member 6 made of SUS is provided at an axial center of the control rod 1 (central tie rod), and an opening portion of a sheath 7 made of SUS having a deep U-shaped cross-section forming an outer periphery of the wing 2 is secured by welding to each projection of this integral type center structural member 6. A plurality of sheath holes 8 and water holes 9 are pierced in the sheath 7, in which two Hf plates 10 which are neutron absorber plates are supported by a load supporting member 12 also serving as a gap (interval) maintaining space, and a water gap 11 (gap through which cooling water flows during use in the reactor) is formed between the two Hf plates 10. The load supporting member 12 has a top-like shape, and the thickness of a gap maintaining portion 12a at the center thereof has a function of spacer. The Hf plates 10 are supported by attaching the Hf plates 10 from both the sides to a support shaft 12b through an attachment hole 13 and causing engagement of the support shaft 12b with a sheath hole (bore) 8, which are secured together by means of welding. When inserting or withdrawing the control rod 1 into or from the reactor core, a percussive force is applied to the sheath 7 upon intermittent driving, or particularly, upon starting driving or decelerating during scram of the reactor. In a long-life type control rod 1, the sheath 7 and the load supporting member 12 made of SUS forming the wing 2 have a thermal expansion coefficient about three times as high as that of the Hf plate 10, which is the neutron absorber formed of a material different from those of the sheath 7 and the load supporting member 12. For example, while SUS has a thermal expansion coefficient of 17.8.times.10.sup.-6 /deg-C., that of Hf is 5.9.times.10.sup.-6 /deg-C. ("Nuclear Reactor Materials Handbook" published by Nikkan Kogyo Shinbun-sha). To avoid inconveniences resulting therefrom, the attachment hole 13 of the Hf plate 10 to be attached to the support shaft 12b of the load supporting member 12 has a diameter larger than that of the support shaft 12b to provide a margin, thereby permitting avoidance of their mutual interference through expansion and contraction in heat cycles during operation of the reactor. In the example shown in FIG. 20, the Hf plate 10 of the control rod 1 having a length L in the inserting direction into the reactor core or the sheath longitudinal direction is longitudinally and equally divided into eight sections. The length 1 of a single Hf plate 10 is therefore about L/8. FIGS. 20 and 21A are described with a scale compressed in the axial direction for convenience of illustration, and FIG. 21B represents the Hf plate substantially similarly to the actual one. Within the wing 2, two Hf plates 10 which are neutron absorber plates arranged opposite to each other form an Hf plate pair 14 which is held by the sheath 7 through four (or three, five or six) load supporting members 12. The attachment hole 13 of the Hf plate and the sheath hole 8 of the sheath 7 have the same pitch 15 size in the sheath longitudinal direction. During inserting or withdrawing operation of the control rod 1, the sheath 7 is subjected, not only to a static load caused by the weight of the Hf plate pair 14 applied through the load supporting member 12 in the stationary state, but also to a dynamic load caused by the relative displacement to the Hf plate pair 14. This load caused by relative displacement becomes an impact load particularly upon intermittent operation or driving starting operation or decelerating operation during a rapid driving in the scram of the reactor. These loads tend to be believed to be shared by the four load supporting members 12 in general and transmitted to the sheath 7. Actually, however, even when a margin is provided taking account of a difference in thermal expansion caused by the difference in thermal expansion coefficient between different materials such as between the attachment hole 13 of the Hf plate and the support shaft 12b of the load supporting member 12, it is conceivable that a single load supporting member bears all the load because of, for example, the manufacturing tolerance. In the worst case, a specific unknown load supporting member 12 receives a large stress, thus causing a local stress to concentrate on the sheath hole 8 at the position where that load supporting member is secured. This is therefore undesirable from the point of view of ensuring soundness of the sheath 7. In the Hf plate 10 of the Hf plate pair 14, as shown in FIG. 20B, the neutron irradiation is larger at a position closer to the insertion leading end and the reactivity value must be increased. The thickness becomes therefore thicker toward the insertion leading end and thinner toward the insertion terminal end. The length in the control rod insertion/withdrawal direction, which is the sheath longitudinal direction of each Hf plate 10, is usually constant, and the thickness of the sheath 7 is also uniform in the control rod insertion/withdrawal direction. A wider water gap corresponds to a higher reactivity value. The sheath 7 should therefore be preferably the thinnest possible. However, since the thickness of the sheath 7 is associated with strength of the sheath 7, an excessive reduction of thickness causes deterioration of mechanical soundness, thus preventing improvement of service life of the control rod 1. More specifically, the distribution in the sheath longitudinal direction of the load acting on the sheath 7 is such that the load is larger toward the insertion leading end because of the thicker Hf plate 10. When designing the thickness of the sheath 7, therefore, it is necessary to sufficiently take account of the weight of the Hf plates 10 serving as the neutron absorber plates and the impact load received upon operation of the control rod 1, as well as mechanical strength with due regards to the manufacturing tolerance. When a horizontal impact is caused by an earthquake or the like, a relatively large stress occurs near the central portion in the longitudinal direction of a long control rod 1. Ensuring a satisfactory mechanical strength for the proximity of the middle is an important task. Therefore, using the control rod 1 for the longest possible period of time contributes to improvement of reliability of the control rod and economic merits of reactor operation. In order to further extend the service life of the long-life type control rod 1, it is necessary to increase mechanical strength having so far formed a restriction as compared with the nuclear life in neutron absorption. SUMMARY OF THE INVENTION A primary object of the present invention is to substantially eliminate defects or drawbacks encountered in the prior art and to provide a control rod for a nuclear reactor, which has a mechanical strength improved through supporting manner of neutron absorbers in a long-life type control rod and has a further longer service life in balance with the nuclear life. Another object of the present invention is to provide a control rod for a nuclear reactor which is improved in mechanical and physical strength, making it possible to use for a long time and achieving reliability and economical merits in the reactor operation. A further object of the present invention is to provide a control rod for a nuclear reactor capable of reducing a load applied per one load supporting member acting on a U-shaped sheath and reducing a local stress applied to the U-shaped sheath to thereby enhance the soundness of the control rod. A still further object of the present invention is to provide a control rod for a nuclear reactor capable of improving a reactivity value as well as improving the soundness of the U-shaped sheath. A still further object of the present invention is to provide a control rod for a nuclear reactor capable of further improving a stress resisting property and the mechanical strength of the U-shaped sheath by separately carrying an integral type neutron absorbing element. A still further object of the present invention is to provide a control rod for a nuclear reactor capable of maintaining the soundness of the load supporting member and the integral type neutron absorbing element to thereby improve the life time and the reliability of the control rod. A still further object of the present invention is to provide a control rod for a nuclear reactor having an improved mechanical and physical strength by increasing and ensuring reliability of a weld portion through abut-welding of the U-shaped sheath against a recessed projection of a central structural member. These and other objects can be achieved according to the present invention, which provides in a general aspect, a control rod for a nuclear reactor comprising a center structural member, a plurality of wings each composed of a sheath member of long plate structure having a U-shaped cross section having an opening which is secured to the center structural member, a front end structural member secured to a front end side of the wing viewed from a wing inserting direction in a reactor core, a terminal end structural member secured to a terminal end side of the wing viewed from the wing inserting direction in the reactor core, a plurality of integral type neutron absorbing elements each having a plate structure accommodated in each of the sheaths in a row in a longitudinal direction thereof and each being formed in plate shape by integrating one or more neutron absorbing plates, and a plurality of load supporting members for supporting weights of the integral type neutron absorbing elements. According to the present invention, the following characteristic features will be provided to the control rod for a nuclear reactor of the structure mentioned above. The length in the sheath longitudinal direction of at least one set of the integral type neutron absorbing elements accommodated in the wing is reduced, the reduced integral type neutron absorbing elements are supported to the U-shaped sheath by the load supporting members to thereby reduce a local load applied to the U-shaped sheath. At least a set of integral type neutron absorbing elements having a relatively large weight are divided into a plurality of sections in a wing width direction, weight of the divided integral type neutron absorbing element sections are supported to the U-shaped sheath respectively by a plurality of load supporting members so that total supporting ability of the plurality of load supporting members which support the divided integral type neutron absorbing element sections surpasses supporting ability of the load supporting members of the integral type neutron absorbing elements not yet divided. The integral type neutron absorbing elements divided into the plurality of sections in the wing width direction is provided with a stiffener containing a neutron absorber to at least one portion of the divided neutron absorbing element and the stiffener is secured to the U-shaped sheath. In the integral type neutron absorbing elements divided into the plurality of sections in the wing width direction, a neutron absorbing ability per unit length in the wing width direction of the integral type neutron absorbing elements located on an outside edge of the wing is increased as compared with that of portions other than the wing outside edge. The integral type neutron absorbing element is formed with an attachment hole into which a support shaft of the load supporting member is inserted with a margin, at least two attachment holes are formed with space in the sheath longitudinal direction respectively on both inside and outside portions in the wing width direction, and a distance between a pair of two load supporting members in the sheath longitudinal direction is made different in accordance with a pitch of the attachment hole in the sheath longitudinal direction being in a range less than a margin of a diameter of the attachment hole exceeding manufacturing tolerance. Two load supporting members paired in the sheath longitudinal direction having the attachment holes of different pitches in the sheath longitudinal direction are disposed inside in the wing width direction. The integral type neutron absorbing element is formed with an attachment hole into which a support shaft of the load supporting member is inserted with a margin, at least two attachment holes are formed with space in the sheath longitudinal direction respectively on both inside and outside portions in the wing width direction, and a pitch in the sheath longitudinal direction of the attachment hole of the integral type neutron absorbing element is different in a range less than a margin of a diameter of the attachment hole exceeding a manufacturing tolerance compared with a pitch in the sheath longitudinal direction between the load supporting members. The attachment hole having a pitch in the sheath longitudinal direction different from a pitch in the sheath longitudinal direction between the load supporting members are formed inside in the wing width direction of the integral type neutron absorbing element. At least a part of the plurality of load supporting members supporting the integral type neutron absorbing element is arranged in a vicinity of a central portion of the integral type neutron absorbing element, and a margin between a support shaft of the load supporting member disposed near the central portion thereof and an attachment hole of the integral type neutron absorbing element into which the support shaft is inserted is made less than a margin between another support shaft of another load supporting member other than that disposed near the central portion of the neutron absorbing element and an attachment hole formed thereto into which the another support shaft is inserted to thereby increase a load supporting ability. The integral type neutron absorbing element is composed of a long-life type neutron absorbing plate by forming a hafnium alloy into a plate shape diluted by diluting hafnium with a diluting agent such as zirconium or titanium, and the integral type neuron absorbing element has a trap structure based on a combination of a plurality of the neutron absorber plates with a water gap serving as a reactor water channel, and the load supporting member is provided with a gap maintaining function between the neutron absorber plates. The integral type neutron absorbing element is formed with an attachment hole into which a support shaft of the load supporting member is inserted with a predetermined margin, the load supporting members are disposed on the control rod inserting front end side and inserting terminal end side of the integral type neutron absorbing element in a manner separated in a wing width direction, and a margin in engagement between either one of support shafts in the wing width direction of the load supporting members and the attachment hole of the integral type neutron absorbing element is made smaller than a margin in engagement between another one support shaft of the load supporting member and the attachment hole of the integral type neutron absorbing element. The integral type neutron absorbing element is provided with an attachment hole to which the load supporting member secured to the sheath is inserted with a predetermined margin, the load supporting members secured to the sheath are attached to the attachment holes of the integral type neutron absorbing elements and also provided with a frictional load supporting member having a gap maintaining function and a frictional resistance function between the neutron absorber plate constituting the integral type neutron absorbing element and the U-shaped sheath. The frictional load supporting member is composed of a projection projecting from an inner surface of the U-shaped sheath, and a recessed portion formed to a surface of the neutron absorber plate and engaged with the projection to thereby imparting the frictional resistance function. The support shafts of the load supporting members secured to the U-shaped sheath are fitted to attachment holes formed to the integral type neutron absorbing element with a predetermined margin, and the support shaft of a specific member of the load supporting members and a corresponding attachment hole of the integral type neutron absorbing element have large diameters in comparison with those of the other load supporting members to improve a load supporting ability. The specific load supporting member is arranged inside in a wing width direction. The plurality of integral type neutron absorbing elements include the neutron absorbers each having a thickness gradually reduced from the leading end side of control rod insertion toward the terminal end side thereof and the length of the neutron absorber is increased toward the insertion terminal end side. The plurality of integral type neutron absorbing elements are reduced in lengths thereof at portions near a central portion of an entire length of the control rod and a load supporting member is arranged also at central portions in a wing width direction and in a sheath longitudinal direction of at least part of the integral type neutron absorbing elements on the control rod insertion end side. The plurality of integral type neutron absorbing elements have substantially uniform lengths throughout an entire length of the control rod and a load supporting member is provided at a central portion in a wing width direction and in a sheath longitudinal direction of the integral type neutron absorbing element up to substantially 2/3 length from the leading end in the control rod inserting direction. The integral type neutron absorbing element is composed of a neutron absorbing plate by forming a hafnium alloy into a plate shape diluted by diluting hafnium with a diluting agent such as zirconium or titanium, the integral type neuron absorbing element has a trap structure based on a combination of a plurality of the neutron absorber plates with a water gap being interposed therebetween, and the opposing neutron absorber plates with the water gap being interposed therebetween are arranged stepwise in a control rod insertion/withdrawal direction. The integral type neutron absorbing element is composed of a neutron absorbing plate by forming a hafnium alloy into a plate shape diluted by diluting hafnium with a diluting agent such as zirconium or titanium, the integral type neuron absorbing element has a trap structure based on a combination of a plurality of said neutron absorber plates with a water gap being interposed therebetween, the neutron absorber plate has a thickness gradually reduced from the leading end side of the control rod insertion toward the terminal end side thereof and the load supporting member to be secured to the U-shaped sheath is fitted to the attachment hole with a predetermined margin, two such load supporting members are each provided at an interval in the sheath longitudinal direction, and an interval maintaining member secured to the U-shaped sheath at central portions in a wing width direction and a sheath longitudinal direction between the load supporting members is mounted to at least one set of the integral type neutron absorbing elements. The neutron absorber plates of at least one set of integral type neutron absorbing elements are made of a hafnium alloy diluted by a diluting agent such as zirconium or titanium and having a hafnium content equal to that in neutron absorber plates of another integral type neutron absorbing element, and the neutron absorber plates made of the hafnium alloy each has an increased thickness to thereby improve mechanical and physical strengths. The integral type neutron absorbing element is composed of a neutron absorbing plate by forming a hafnium alloy into a plate shape diluted by diluting hafnium with a diluting agent such as zirconium or titanium, the integral type neuron absorbing element is formed with a water gap in a control rod insertion/withdrawal direction in the U-shaped sheath and has substantially a box-shaped sectional shape in a direction perpendicular to the control rod insertion/withdrawal direction. The center structural member is formed with a recessed projection having a thickness equal to that of the wing and extending in a longitudinal direction thereof on a side to be secured to the U-shaped sheath. According to the present invention of the various aspects mentioned above, there are achieved the following functions and advantageous effects. In one aspect, there is provided a control rod for a nuclear reactor constructed by securing a leading end structural member to a leading end, and a terminal structural member to a terminal end in the core inserting direction by means of a long sheath having a U-shaped cross-section, securing or integrating flat with a tolerance of slight sliding one or more sheet-shaped neutron absorbing plates divided into a plurality of sections in the sheath longitudinal direction, housing the integral type neutron absorbing elements into the sheath, forming a plurality of wings so as to hold the weight of the individual integral type neutron absorbing elements through a plurality of load supporting members by the sheath, and securing a U-shaped opening of each of the plurality of wings to an opening side structural member formed by an integral type center structural member of an independent type structural member, and the length in the sheath longitudinal direction of at least one set of integral type neutron absorbing elements, of which the weight becomes relatively large, is reduced to alleviate a local load exerted on the sheath. According to the above aspect of the present invention, by reducing the weight by reducing the length of the integral type neutron absorbing elements which are relatively heavy at the insertion leading end into the reactor core, the load per load supporting member is reduced, thereby securing the load supporting member and reducing local stress at the position where the load supporting member is secured in the sheath holding the load of the integral type neutron absorbing elements. At least a set of integral type neutron absorbing elements having a relatively large weight are divided into a plurality of sections in a direction at right angles to the longitudinal direction of the sheath, so that the total supporting ability of the plurality of load supporting members which support the divided integral type neutron absorbing elements surpasses the supporting ability of the load supporting members of the integral type neutron absorbing elements not as yet divided. Furthermore, because the load per load supporting member is reduced by the reduction of weight of the integral type neutron absorbing elements, local stress on the sheath holding the load of the integral type neutron absorbing elements by securing the load supporting member is reduced. According to the present invention, there is provided a control rod for a nuclear reactor, in the integral type neutron absorbing elements divided into the plurality of sections in a direction normal to the longitudinal direction of the sheath, a stiffener containing a neutron absorber is arranged at least at a place in the integral type neutron absorbing element sections resulting from division of the integral type neutron absorbing elements into the plurality of sections, and the stiffener is secured to the sheath. Accordingly, mechanical and physical strength of the sheath is improved because the load applied per load supporting member is reduced, and further, since a long-life type neutron absorber is housed in the stiffener, there occurs almost no decrease in the reactivity value of the control rod. According to the present invention, in the integral type neutron absorbing elements divided into the plurality of sections in a wing width direction, the neutron absorbing ability per unit length in the wing width direction of the integral type neutron absorbing elements located on the outer edge of the wing is increased as compared with that of portions other than the wing outer edge. Therefore, the reduction of the load acting on the load supporting member leads to an improved soundness of the sheath. When using integral type neutron absorbing elements of the same weight, the reactivity value of the control rod can be improved with an extended nuclear life. When it is not necessary to improve the reactivity value or the nuclear life, the weight is reduced by reducing the quantity of integral type nuclear absorbing elements, thereby further improving soundness of the sheath supporting the same. According to the present invention, the integral type neutron absorbing elements are attached to the support shaft and the attachment hole of the load supporting member secured to the sheath in the attachment hole with a predetermined margin, at least two such load supporting members each are provided at an interval in the sheath longitudinal direction, and the distance between predetermined two load supporting members forming a pair in the sheath longitudinal direction varies within a range in which the pitch between a pair of attachment holes in the integral type neutron absorbing elements is over a manufacturing tolerance and under the margin of the attachment hole. When the integral type neutron absorbing elements move relative to the sheath, any one of the two specified load supporting members forming a pair in the sheath longitudinal direction bears the load when inserting or when withdrawing. The two load supporting members are specified depending upon the direction of insertion or withdrawal, thus certainly sharing the load. The load per each load supporting member is therefore reduced to a half, and this improves, together with the load supporting members, stress resistance and mechanical strength of the sheath. According to the present invention, the two load supporting members forming a pair in the sheath longitudinal direction, different from the pitch between a pair of the attachment holes are closer to the structural member for securing the wing. In the sheath of the wing, the side secured to the structural member has a higher mechanical strength than that on the opposite side. By causing the load supporting member closest to this structural member to bear the load of the integral type neutron absorbing elements, the stress resistance and the mechanical strength of the sheath securing this load supporting member can be improved. According to the present invention, the integral type neutron absorbing elements are attached to the support shaft and the attachment hole of the load supporting member secured to the sheath in the attachment hole with a predetermined margin, at least two such load supporting members each are provided at an interval in the sheath longitudinal direction, and as compared with the pitch between a pair of the load supporting members, the distance between the two prescribed load supporting members forming a pair in the sheath longitudinal direction varies within a range in which the pitch between a pair of attachment holes in the integral type neutron absorbing elements is over a manufacturing tolerance and under the margin of the attachment hole. When the integral type neutron absorbing elements move relative to the sheath, any one of the two prescribed load supporting members forming a pair in the sheath longitudinal direction bears the load when inserting or when withdrawing. The two load supporting members are specified depending upon the direction of insertion or withdrawal, thus certainly sharing the load. The load per each member is therefore reduced to a half, and this improves, together with the load supporting members, the stress resistance and the mechanical strength of the sheath. According to the present invention, the attachment holes of the two integral type neutron absorbing elements forming a pair in the sheath longitudinal direction are closer to the structural member securing the wing. In the sheath of the wing, the side secured to the center structural member has a higher mechanical strength than that on the opposite side. By Causing the load supporting member closest to this structural member to bear the load of the integral type neutron absorbing elements, the stress resistance and the mechanical strength of the sheath securing this load supporting member can be improved. According to the present invention, a part of the plurality of load supporting members are provided at least in the proximity of the center of the integral type neutron absorbing elements, and the margin between the support shaft of the load supporting member near the central portion and the attachment hole of the integral type neutron absorbing element is reduced to under the margin between the other load supporting members and the attachment hole, thereby increasing the load supporting ability. The plurality of load supporting members provided closer to the central portion of heavy integral type neutron absorbing element shares the impact load and the like by a long engagement with the integral type neutron absorbing element, thereby improving the load resistance of the sheath and ensuring intervals between the integral type neutron absorbing elements. According to the present invention, the integral type neutron absorbing element has a trap structure based on a combination of a plurality of neutron absorber plates prepared from hafnium metal or by forming a hafnium alloy made by diluting hafnium with zirconium or titanium into plates with a water gap serving as a reactor water channel, and an interval (gap) maintaining function between the neutron absorber plates to the load supporting members. By realizing the trap structure by using neutron absorber plates made of hafnium metal or a hafnium alloy as integral type neutron absorbing elements and arranging them opposite to each other with a water gap therebetween, while maintaining the interval near the central portion, it is possible to eliminate a deflection toward inside of the neutron absorber plates, prevent a decrease in reactivity value and improve mechanical strength of the sheath. According to the present invention, the load supporting members secured to the sheath are attached to the attachment hole of the integral type neutron absorbing element with a predetermined margin, the load supporting members are provided at an interval in a direction normal to the sheath longitudinal direction on the leading end side and on the terminal end side of insertion of the control rod in the integral type neutron absorbing element, and a margin between the support shaft of any of the load supporting members and the attachment hole of the integral type neutron absorbing element is reduced to under the margin between the other load supporting member and the attachment hole. The load on the integral type neutron absorbing elements is certainly borne by sharing by the plurality of load supporting members in any of the inserting and withdrawing directions of the control rod, thus alleviating a local impact load of the sheath securing the load supporting members and improving the soundness of the sheath. According to the present invention, the load supporting members secured to the sheath are attached to the attachment holes of the integral type neutron absorbing elements, and there is provided a frictional load supporting member having an interval maintaining function and a frictional resistance function between the neutron absorber plate of the integral type neutron absorbing element and the sheath. Relative motion of the sheath and the neutron absorber plates of the integral type neutron absorbing elements along with operation of the control rod inhibits an impact load through frictional resistance between the sheath and the neutron absorber plates given by the frictional load supporting members. As a result, the burden on the load supporting members is alleviated and the soundness of the sheath is improved. According to the present invention, the frictional resistance function of the frictional load supporting member causes engagement of a dimpling projecting into the inner surface of the sheath and a recess formed on the surface of the neutron absorber plate. By realizing the trap structure by using neutron absorber plates made of hafnium metal or a hafnium alloy as integral type neutron absorbing elements and arranging them opposite to each other with a water gap therebetween, while maintaining an interval near the central portion, it is possible to eliminate a deflection toward inside of the neutron absorber plates, prevent a decrease in the reactivity value and improve the mechanical strength of the sheath. According to the present invention, the load supporting members secured to the sheath are attached to the attachment hole of the integral type neutron absorbing element with a prescribed margin, the load supporting members are provided at an interval in a direction normal to the sheath longitudinal direction on the leading end side and on the terminal end side of insertion of the control rod in the integral type neutron absorbing element, and a margin between the support shaft of any of the load supporting members and the attachment hole of the integral type neutron absorbing element is reduced under the margin between the other load supporting member and the attachment hole. The load on the integral type neutron absorbing elements is certainly borne by sharing by the plurality of load supporting members in any of the inserting and withdrawing directions of the control rod,thus alleviating a local impact load of the sheath securing the load supporting members and improving the soundness of the sheath. According to the present invention, the load supporting members secured to the sheath are attached to the attachment holes of the integral type neutron absorbing elements, and there is provided a frictional load supporting member having an interval maintaining function and a frictional resistance function between the neutron absorber plate of the integral type neutron absorbing element and the sheath. The relative motion of the sheath and the neutron absorber plates of the integral type neutron absorbing elements along with operation of the control rod inhibits an impact load through frictional resistance between the sheath and the neutron absorber plates given by the frictional load supporting members. As a result, the burden on the load supporting members is alleviated and the soundness of the sheath is improved. According to the present invention, the frictional resistance function of the frictional load supporting member causes engagement of a dimpling projecting into the inner surface of the sheath and a recess formed on the surface of the neutron absorber plate. By achieving engagement of the dimpling of the t sheath and the recess of the neutron absorber plate and holding this state with the frictional load supporting member, frictional resistance occurs between them. According to the present invention, the support shaft of the load supporting member secured to the sheath is attached to the attachment hole of the integral type neutron absorbing element with a predetermined margin, and the support shaft of the load supporting member and the attachment hole of the integral type neutron absorbing element have large diameters at a specific load supporting member position from among the load supporting members, thereby improving the load supporting ability. Because of the large diameters of the attachment hole, the support shaft of the load supporting member and the sheath hole which support the impact load and the weight of the integral type neutron absorbing elements, the load bearing ability is increased and the soundness of the sheath and the like is improved. According to the present invention, the specific load supporting member at the position of which the supporting shaft of the load supporting member and the attachment hole of the integral type neutron absorbing element have larger diameters than the others is closer to the center structural member securing the wing. In the sheath of the wing, the side secured to the center structural member has a higher mechanical strength than that on the opposite side. By causing the load supporting member closest to the center structural member to bear the load of the integral type neutron absorbing elements, the stress resistance and the mechanical strength of the sheath securing this load supporting member can be improved. According to the present invention, the thickness of the neutron absorber is gradually reduced from the leading end side of control rod insertion toward the terminal end side, and the length of the neutron absorber is increased toward the terminal end side. Since the weight is substantially uniform for all the neutron absorber plates, the impact load borne by the individual load supporting members are substantially equalized to the load stress in the sheath, thereby improving the soundness of the sheath and the like. According to the present invention, the length of the plurality of integral type neutron absorbing elements is reduced near the central portion of the entire length of the control rod, and a load supporting section is provided also at the central portion in the wing width direction and in the longitudinal direction of at least a part of the integral type neutron absorbing elements. Therefore, seismic resistance is improved by increasing the attachment density of the load supporting members through reduction of the length of the integral type neutron absorbing elements at the central portion of the entire length of the control rod receiving a large stress in an earthquake. Addition of the load supporting members to the relatively heavy integral type neutron absorbing elements reduces the supported load per load supporting member, thereby improving the soundness of the sheath. According to the present invention, the plurality of integral type neutron absorbing elements have substantially uniform lengths throughout the entire length of the control rod, and a load supporting member is provided also at the central portion in the width direction and in the longitudinal direction of the integral type neutron absorbing element up to about 2/3 length from the leading end in the control rod inserting direction. The addition of load supporting members to the relatively heavy integral type neutron absorbing elements and to the portions receiving a large stress in an earthquake reduces the supported load per load supporting member, thereby improving the soundness of the sheath. According to the present invention, the integral type neutron absorbing element has a trap structure based on a combination of a plurality of neutron absorber plates prepared from hafnium metal or by forming a hafnium alloy made by diluting hafnium with zirconium or titanium into plates with a water gap therebetween, and neutron absorber plates having the water gap is arranged stepwise in the control rod inserting direction. The integral type neutron absorbing elements in the sheath comprise neutron absorber plates made of hafnium metal or a hafnium alloy, arranged stepwise with the water gap. The strength in the horizontal direction is therefore reinforced and the strength of the sheath is improved. In the integral type neutron absorbing elements, there is no longitudinal gap between the neutron absorber plates, with no crossing neutrons, thus permitting improvement of reactivity value of the control rod. According to the present invention, the integral type neutron absorbing element has a trap structure based on a combination of a plurality of neutron absorber plates prepared from hafnium metal or by forming a hafnium alloy made by diluting hafnium with zirconium or titanium into plates with a water gap and the thickness of the neutron absorber plate is gradually reduced from the leading end side of the control rod insertion toward the terminal end side thereof, the load supporting member secured to the sheath is attached to the attachment hole with a predetermined margin, two such load supporting members are each provided at an interval in the sheath longitudinal direction, and an interval maintaining member secured between the sheaths of both the surfaces at the central portion in the longitudinal direction and in the width direction between the load supporting members is provided on each of at least on set of integral type neutron absorbing elements. Since the interval maintaining member is secured between two oppositely arranged neutron absorber plates at the central portion from the load supporting member, the neutron absorbing plates become difficult to be bent, and the sheath strength is reinforced, thus improving the soundness of the sheath. An appropriate gap is kept between the neutron absorber plates arranged opposite to each other, thereby preventing a decrease in reactivity value of the control rod. According to the present invention, at least one set of integral type neutron absorbing elements, the neutron absorber plates are made of a hafnium alloy comprising hafnium and a metal having a specific gravity smaller than that of hafnium which forms an alloy with hafnium such as zirconium or titanium, at a content of hafnium equal to that in hafnium metal, and therefore, the mechanical strength is improved by increasing thickness. A neutron absorber plate in an integral type neutron absorbing element made of a hafnium alloy has an increased thickness as compared with hafnium metal having the same hafnium content. This increases strength and reinforces the sheath strength, resulting in the improved soundness of the sheath. According to the present invention, the integral type neutron absorbing elements comprise neutron absorber plates formed from hafnium metal or a hafnium alloy prepared by diluting hafnium with zirconium or titanium into a plate shape, the integral type neutron absorbing elements form water passages in the control rod inserting direction in the sheath and the cross-section in a direction at an angle normal to the insertion/withdrawal direction forms substantially a box shape. Since the integral type neutron absorbing element has a substantially box-shaped cross-section, there is an increase in toughness, leading to a higher mechanical strength, and an increased sheath strength improves the soundness of the sheath. The water passages are provided on both the sides of the integral type neutron absorbing element in the sheath. According to the present invention, a projection having the same thickness as that of the sheath is provided on a center structural member (tie rod) and the projection having a thickness equal to the wing thickness is welded together with the sheath. Welding of plates having the same thickness leads to a uniform heat input to both the plates in the welding, thus reducing welding defects such as insufficient penetration. |
claims | 1. A nuclear reactor, comprising: a housing with a reflector, forming a reactor core including an inner space; first process channels, located in the reactor core, designed for coolant circulation; second process channels, located in the reactor core, designed for placement of control and protection system components; a plurality of fuel rod arrays; a first coolant loop; wherein:the first coolant loop comprises a supply chamber including a bottom and a discharge chamber separated from the supply chamber by a partition; the first process channels are designed as bayonet tubes, each said bayonet tube includes an external tube and an internal tube, each said external tube is attached to the bottom of the supply chamber, and each said internal tube is attached to the partition; each fuel rod array of said plurality of fuel rod arrays is installed on a suspender, inside of each corresponding said internal tube, wherein the suspender is attached to an upper part of the discharge chamber; the second process channels are isolated from the supply chamber and the discharge chamber; and the inner space of the reactor core is filled with medium or material transparent for neutrons. 2. The nuclear reactor according to claim 1, wherein: the reflector comprising a side reflector designed as a pack of rings, an upper reflector and a lower reflector. 3. The nuclear reactor according to claim 1, wherein: the inner space is filled with a zirconium alloy. 4. The nuclear reactor according to claim 1, wherein: said control and protection system components include control and protection system controls located at an upper part of the discharge chamber. 5. The nuclear reactor according to claim 1, wherein: said control and protection system components include emergency protection absorbing rods, compensating rods, and absorbing control rods. 6. The nuclear reactor according to claim 5, wherein: the compensating rods and the emergency protection absorbing rods include an absorber consisting of B4C enriched to 80% for 10B. 7. The nuclear reactor according to claim 5, wherein: the control rods include an absorber consisting of B4C enriched to 20% for 10B. 8. The nuclear reactor according to claim 1, wherein: said plurality of fuel rod arrays include a part of fuel rods filled with Gd2O3 burnable absorber. 9. The nuclear reactor according to claim 1, wherein: said plurality of fuel rod arrays include a part of fuel rods filled with Er burnable absorber. 10. The nuclear reactor according to claim 1, wherein: said plurality of fuel rod arrays include a first number of fuel rods filled with Gd2O3 burnable absorber and a second number of fuel rods filled with Er burnable absorber. |
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053135067 | description | DESCRIPTION OF THE PREFERRED EMBODIMENT Referring to FIG. 1, a typical prior art fuel bundle B is shown. The fuel bundle B includes a lower tie plate 14, and upper tie plate 16 and a plurality of fuel rods F in a matrix extending between the respective tie plates. The tie plates 14, 16 as well as the fuel rods F are surrounded by a channel C, which channel C is broken away so that a single spacer S can be seen. Spacer S is the subject of this invention. More particularly, a spacer S of the ferrule type is the subject of this patent application. FIG. 2 is an illustration of a particularly advantageous spring 12. This spring is taken from Johansson et. al. U.S. Pat. application Ser. No. 07/623,828 entitled Self Locating Springs for Ferrule Spacer filed Dec. 6, 1990 at FIG. 17. The content of this patent application is herein incorporated by reference. This loop spring, illustrated in the attached specification at FIG. 2, includes first and second vertical legs 14 and 16 containing rod contacting portions 15, 17. One vertical leg at one rod contacting portion extends interior of a first ferrule of a ferrule spacer to bias a fuel rod within that ferrule. The other vertical leg at the other rod contacting portion extends interior of an adjacent second ferrule of a ferrule spacer to bias the adjacent fuel rod within that adjacent ferrule. This disclosed spring has a shortened construction. This shortened construction is provided by two horizontal loops 20, 24; these horizontal loops being positioned one loop 20 at the top of the spring and the remaining loop 24 at the bottom of the spring. The first and top horizontal loop is connected at its central portion to the top of each of the spring legs 14, 16. The second and bottom horizontal loop is connected at its central portion to the bottom of each of the spring legs 14, 16. Thus, the spring when compressed flexes over the vertical length of the spring legs 14, 16 as well as the horizontal length of the top and bottom loops 20, 24. The respective top and bottom horizontal loops 20, 24 are provided with expanded loop ends 21, 22 in loop 20 and 25, 26 in loop 24. This expansion is labeled in an exemplary fashion at expanded sections 28, 30 in expanded end 25 of loop 24. A spring having a reduced vertical dimension with a superior deflection range results. As will hereinafter become more apparent, I prefer to trap this spring at the expanded ends 25 between opposing ferrules for the construction of the preferred spacer of this invention. Referring to FIGS. 3A and 3B the first embodiment of a ferrule pair 40 can be seen and understood. With reference to FIG. 3B, ferrule 42 includes solid section 42A and apertured section 42B having aperture 60 therein. Likes wise ferrule 44 includes solid section 44A and apertured section 44B. It will be appreciated that in each of the apertured sections, four apertures 60 appear. Appropriate stops 70 are configured interior of the ferrules. It will be further understood that between each ferrule pair 42, 44, there is defined a spring capturing window. Spring 12 (See FIG. 2) is first place within the aperture 62 of partially confronted ferrules and thereafter confrontation occurs. There results the composite construction which is the essential building element of the spacer construction of this invention. Referring to FIG. 4, a completed spacer S is shown using the ferrules of FIGS. 3A and 3B. This spacer S defines intervals for water rods W1 and W2. As the bracing of these water rods W1 and W2, and the exterior band are conventional, they are not shown. Referring to FIGS. 3A and 3B, the preferred embodiment of a ferrule pair 40 can be seen and understood. Referring to FIG. 3B, first ferrule section 42 is shown confronted to second ferrule section 44. Examining FIG. 3B further, it can be seen that a centerline 46 divides each ferrule of the ferrule pair 42, 44 into respective halves. Ferrule 42 has halves 42A and 42B; ferrule 44 has halves 44A and 44B. Provision must be made for the confrontation of the respective ferrule halves so as to define only one ferrule wall even though two ferrules are confronted. Accordingly, ferrule halves 42B and 44B have four portions of their respective walls milled away. Specifically, a milling tool is utilized which has two characteristics. First, the milling tool has a diameter slightly exceeding the diameter of the adjacent ferrules. Secondly, the milling tool when cutting away the wall is centered just as the adjacent ferrule would be centered. The result is milled intervals 60. Four such intervals are placed in ferrule halves 42B and four such intervals are milled in ferrule halves 44B. In order to accommodate the subsequent fit, it will be observed that the milling process exceeds in vertical dimension slightly the half dimension of the spacer halves 42B and 44B. The fit between the respective ferrule halves is as shown in FIGS. 3A and 3B. Referring to FIG. 3B, it can be seen that ferrule half 42A fits adjacent ferrule half 44B; likewise it can be seen that ferrule half 42B fits adjacent ferrule half 44A. As is shown in the plan view of FIG. 3A, with such an arrangement, the contiguous and tangent portions of ferrule halves 42A and 44B define only one ferrule wall between adjacent fuel rods F1, F2. That is, as best seen in FIG. 3A, the ferrule half 42a with walls present includes a wall portion which extends within the slot 60 of the ferrule half 44b and consequently only a single wall lies between adjacent fuel rods along a line extending parallel to the axes of the ferrules and along a tangency between the ferrule pair 40. The rest of the construction is conventional. A ferrule spring 12 is trapped between ferrule aperture 62 in ferrule 42 and ferrule aperture 64 in ferrule 44. As has been illustrated in Johansson et al. U.S. Pat. application Ser. No. 07/623,828 entitled Self Locating Springs for Ferrule Spacer filed Dec. 6, 1990, spring 12 is captured at the respective loops interior of the confronted ferrules. As is further conventional, the respective ferrule halves 42A, 42B, 44A and 44B define respective stops 70 into which fuel rods F1 and F2 are biased by spring 12. Turning to FIG. 4, a spacer S constructed of the ferrules is illustrated. In the plan view of FIG. 6, the orientation of two water rods W1 and W2 is shown. As bracing of the water rods at the spacer S is conventional, it is not shown. This invention can be applied to spacers. S having octagon shaped sections. Such is shown in FIGS. 5, 6A, 6B and 7. Referring to FIG. 5, a Zircaloy metal sheet is shown immediately before being bent into an octagon shape. As before, the cell 80 is divided into halves 80A and 80B. In the upper half 80B of spacer 80, four walls are removed at 82, 84, 86, and 88. These removals slightly intrude upon half 80A of the ferrule cell metal sheet. Conventional apertures 90A and 90B are configured for trapping spring 12. Referring to FIGS. 6A and 6B, two respective cells 80 are assembled. Assembly is by stamping and bending using conventional manufacturing techniques. Consequently, a detailed explanation of the cell formation of the octagon shaped cell will not be offered here. Referring to FIG. 6B, cells 80 are shown juxtaposed with sections 80A of one cell 80 juxtaposed to section 80B of an adjacent cell 80. Spring 12 is trapped between the cells at apertures formed by 90A and 90B between the respective cells. Referring to FIG. 6A, fuel rods F1 and F2 are shown biased into conventionally constructed stops 70 by spring 12. It will be observed that only one cell wall is present between fuel rods F1 and F2 Referring to FIG. 7, a completed spacer S is shown assembled. This spacer S defines intervals for water rods W1 and W2. As the bracing of these water rods W1 and W2 at the spacer and the exterior band are conventional, they are not shown. The reader will appreciate that the construction technique and process here shown will find applicability in arrays other than the 10 by 10 arrays here illustrated. Further, we have shown one spacer cell half having all of the walls removed. Other combinations of wall removal may be utilized. We prefer the illustrated configuration as providing a reversible cell assembly with a minimum inventory of required parts for assembly of the spacer. |
050826034 | summary | BACKGROUND OF THE INVENTION The present invention relates to a method of treatment of a high-level radioactive waste generated, for example, from reprocessing of spent nuclear fuels. More particularly, the present invention is concerned with a method for treating a high-level radioactive waste which comprises adding a suitable amount of boron or a boron compound to a calcined material of the high-level radioactive waste, treating the resultant mixture at a high temperature to alloy platinum group elements contained in the waste with boron, separating and recovering the resultant alloys, and solidifying residual oxides as a solid waste of a high degree of volume reduction. A high-level radioactive waste generated from reprocessing of spent fuels by purex process is stored in the form of a nitric acid solution containing fission products. This liquid waste is solidified in the future through inclusion in a medium such as glass. Besides glass, many materials such as synthetic rock and the like have been studied as the medium. The concentration of the fission products in the medium is limited to about 10% by weight from the viewpoint of problems such as the solubility of the fission products into the medium, chemical durability (leaching rate in water), and removal of decay heat. The volume of the solidified waste should be as small as possible for the purpose of decreasing the cost of storage and disposal thereof. Although the fission products content of the solidified waste must be increased for this purpose, it is difficult at the present time due to the reasons described above. Meanwhile, the high-level radioactive waste contains platinum group elements (Ru, Pd and Rh) which are useful but poor in natural resources. Various attempts have been made to recover these elements, and examples of the known method include: (1) a solvent extraction method wherein these elements are separated from a nitric acid solution of the highd-level radioactive waste by using a phosphoric ester; PA1 (2) a lead extraction method wherein the high-level radioactive waste is vitrified and these elements are extracted from the vitrified waste by using molten lead; and PA1 (3) an ion-exchange method wherein these elements are separated by treating a nitric acid solution of the radioactive waste with an ion-exchange resin. PA1 (1) In the solvent extraction method, the phosphoric ester becomes a secondary waste which is different from the solvent for extraction in the reprocessing, i.e. TBP (tributyl phosphite). This makes it necessary to conduct research and development on a processing method and construction of a processing plant different from those of the waste TBP. The cost necessary for this purpose is very high and causes the cost of the recovery of the platinum group elements to be increased over that of the commercially available platinum group elements, so that the conventional solvent extraction method does not economically pay. PA1 (2) The lead extraction method is advantageous in that lead which becomes a solid waste as it is is used as the extractant. In this method, however, in order to enhance the extraction efficiency, it is necessary to use a low-viscosity glass having a composition different from that of the glass used for the vitrification of the high-level radioactive waste. Further, lead and the platinum group elements should be re-separated, thus rendering this method difficultly applicable to practical use. PA1 (3) The ion exchange method has a problem of safety because a flammable substance is formed when the ion-exchange resin comes into contact with nitric acid. However, these prior art methods of recovering platinum group elements have the following drawbacks. A large amount of a secondary waste occurs in any of the above-described prior art methods, so that a treatment for remarkably reducing the volume of the high-level radioactive waste cannot be accomplished. SUMMARY OF THE INVENTION It is therefore an object of the present invention to provide a novel and improved method for treatment a high-level radioactive waste which can eliminate the above-described drawbacks of the prior art methods and easily recover valuable platinum group elements contained in the radioactive waste. It is another object of the present invention to provide a novel and improved method of treatment of a high-level radioactive waste which does not generate a large amount of a secondary waste and can obtain a highly volume-reduced high-level radioactive solid waste. According to the present invention, in order to accomplish the above-described objects, there is provided a method of treatment of a high-level radioactive waste comprising adding boron or a boron compound to a calcined material of the radioactive waste in an amount of 0.5 to 10% by weight in terms of boron as a simple substance, heating the resultant mixture at a temperature of about 1000.degree. C. or above under a reduction condition to melt the mixture and to alloy platinum group elements present in the calcined material with boron, recovering a layer of the resultant platinum group element alloys from a layer of residual oxides through sedimentation, and solidifying the layer of the residual oxides to form a volume-reduced high-level radioactive solidified waste. The present invention as described above has been made on the basis of our finding that the addition of a suitable amount of boron or a boron compound in the heat melting of the calcined material of the high-level radioactive waste enables a melting treatment temperature to be remarkably lowered because boron alloys with the platinum group elements to form platinum group element alloys which melt at a temperature of about 2000.degree. C. or below. |
claims | 1. A charged particle radiation device having a charged particle source, an extraction electrode for extracting charged particles from the charged particle source, a sample holding means for holding a sample to be irradiated with charged particles extracted by the extraction electrode, a charged particle optical system for irradiating the sample held on the sample holding means with the charged particles that are extracted, a first evacuation means for evacuating a first vacuum chamber in which the charged particle source is arranged, and a second vacuum chamber, independent of the first vacuum chamber, for evacuating a second vacuum chamber connected to the first vacuum chamber, the charged particle radiation device further having:a shielding electrode of a cylindrical structure that is so arranged as to surround the charged particle source and shield against the procession of back scattered charged particles from the extraction electrode, wherein:the top end and the bottom end of the cylinder of the cylindrically structured shielding electrode are open to the inside of the first vacuum chamber. 2. The charged particle radiation device according to claim 1, wherein:the shielding electrode is arranged farther away than the shortest distance between the tip of the charged particle source and the extraction electrode. 3. The charged particle radiation device according to claim 1, wherein:the extraction electrode has a protrusion smaller than the shortest distance between the tip of the charged particle source and the extraction electrode. 4. The charged particle radiation device according to claim 1, wherein:the extraction electrode is a convex extraction electrode. 5. The charged particle radiation device according to claim 1, wherein:the height of the tip of the charged particle source is arranged within the range in the height direction of the first evacuation means. 6. The charged particle radiation device according to claim 1, wherein:the first evacuation means is an entrapment type pump. 7. The charged particle radiation device according to claim 6, wherein:the entrapment type pump is a non-evaporative getter pump or a titanium sublimation pump. 8. A charged particle radiation device having a charged particle source, an extraction electrode for extracting charged particles from the charged particle source, a sample holding means for holding a sample to be irradiated with charged particles extracted by the extraction electrode, a charged particle optical system for irradiating the sample held in the sample holding means with the charged particles that are extracted, a first evacuation means for evacuating a first vacuum chamber in which the charged particle source is arranged, and a second vacuum chamber, independent of the first vacuum chamber, for evacuating a second vacuum chamber connected to the first vacuum chamber, the charged particle radiation device further having:a shielding electrode of a cylindrical structure that is so arranged as to surround the charged particle source and shield against the procession of back scattered charged particles from the extraction electrode, wherein:at least one opening is provided in the side face of the cylinder of the cylindrically structured shielding electrode. 9. The charged particle radiation device according to claim 8, wherein:the shielding electrode is arranged farther away than the shortest distance between the tip of the charged particle source and the extraction electrode. 10. The charged particle radiation device according to claim 8, wherein:the plurality of openings provided in the side of the cylinder are provided with protrusions directed outside the cylinder. 11. The charged particle radiation device according to claim 8, wherein:the shielding electrodes are arranged in a plurality in a telescopic way, andthe opening provided in each shielding electrode is arranged in a position out of alignment with others not to superpose correspondingly to the loci of the back scattered charged particles. 12. The charged particle radiation device according to claim 8, wherein:the side of the shielding electrode has an inclination relative to the center axis thereof. 13. The charged particle radiation device according to claim 8, wherein:the shielding electrode has a mesh structure. 14. The charged particle radiation device according to claim 8, wherein:power sources capable of independently applying voltages to the shielding electrode, the charged particle source and the extraction electrode are installed. 15. The charged particle radiation device according to claim 8, wherein:the power source to apply a voltage to the shielding electrode is common to the power source to apply a voltage to the charged particle source. 16. The charged particle radiation device according to claim 8, wherein:the power source to apply a voltage to the shielding electrode is common to the power source to apply a voltage to the extraction electrode. 17. The charged particle radiation device according to claim 8, further having:an electrode that restrains the back scattered charged particles toward above the shielding electrode. 18. The charged particle radiation device according to claim 8, further having:a directional energy source that gives heat or electric fields to the tip of the charged particle source. 19. The charged particle radiation device according to claim 18, wherein:the directional energy source is a semiconductor laser. 20. The charged particle radiation device according to claim 8, further having:a means to heat the shielding electrode. |
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claims | 1. A method of inspecting a substrate using electrons, the method comprising:obtaining multiple images of a region of the substrate using mirror-mode electron-beam imaging at a range of voltage differences between an electron source and a substrate, each said image corresponding to one said voltage difference;storing image data corresponding to the multiple voltage differences; andcalculating a measure of variation of an imaged aspect of a feature in the region with respect to the voltage difference between the electron source and the substrate,wherein the method provides sub-volt voltage contrast between said images without needing to use an energy-filter with sub-volt resolution. 2. The method of claim 1, wherein the imaged aspect of the feature comprises an intensity of the feature. 3. The method of claim 1, wherein the imaged aspect of the feature comprises an apparent size of the feature. 4. The method of claim 1, further comprising determining a process-related parameter based at least in part on the calculated measure of variation of the imaged aspect of the feature with respect to the voltage difference. 5. The method of claim 4, further comprising obtaining and storing image data from multiple locations on the substrate. 6. The method of claim 5, further comprising building a map showing variation of the process-related parameter over the substrate. 7. The method of claim 1, wherein the voltage difference is varied by changing a voltage applied to a stage holding the substrate. 8. The method of claim 1, wherein the voltage difference is varied by changing a voltage at the electron source. 9. The method of claim 1, wherein the substrate comprises a semiconductor wafer, and wherein the method is applied in-line during the fabrication of integrated circuits on the semiconductor wafer. 10. The method of claim 1, further comprising providing a higher-energy electron beam for substrate charge control. 11. The method of claim 1, further comprising illuminating the region with an ultraviolet light beam for substrate charge control. 12. The method of claim 1, wherein the region of the substrate being imaged includes test structures positioned in between integrated circuit dies on a semiconductor wafer. 13. An apparatus comprising:an electron beam instrument configured for obtaining multiple images of a region of the substrate using mirror-mode electron beam imaging at a range of voltage differences between an electron source and a substrate, each said image corresponding to one said voltage difference;a data processing system configured to store image data corresponding to the multiple voltage differences; andprocessor-executable code configured to calculating a measure of variation of an imaged aspect of a feature in the region with respect to the voltage difference between the electron source and the substrate,wherein the apparatus provides sub-volt voltage contrast between said images without needing to use an energy filter with sub-volt resolution. 14. The apparatus of claim 13, further comprising processor-executable code configured to determine a process-related parameter based at least in part on the calculated measure of variation of the imaged aspect of the feature with respect to the voltage difference. 15. The apparatus of claim 14, further comprising processor-executable code configured to obtain and store image data from multiple locations on the substrate. 16. The apparatus of claim 15, further comprising processor-executable code configured to build a map showing variation of the process-related parameter over the substrate. 17. The apparatus of claim 13, wherein the electron beam instrument utilizes a second electron source to provide an electron beam for substrate charge control. 18. The apparatus of claim 13, further comprising a UV light source configured to illuminate the region for surface charge control. 19. An apparatus for inspecting a substrate using a projection electron beam system, the apparatus comprising:an illumination subsystem configured to generate an incident electron beam;an objective subsystem configured to receive the incident electron beam, to focus the incident electron beam onto a region of the substrate, and to retrieve a scattered beam from the substrate;a projection subsystem configured to receive the scattered beam and to project the scattered beam onto a detector so as to detect electron images of the region; anda beam separator coupled to and interconnecting the illumination subsystem, the objective subsystem, and the projection subsystem, wherein the beam separator is configured to receive the incident beam from the illumination subsystem, bend the incident electron beam towards the objective subsystem, receive the scattered beam from the objective subsystem, and bend the scattered beam towards the projection subsystem, anda control system configured to vary a potential difference between a source of the incident electron beam and the substrate,wherein the electron images are obtained of using mirror-mode electron-beam imaging at a range potential differences between the source and the substrate, each said image corresponding to one said potential difference, andfurther wherein the apparatus provides sub-volt voltage contrast between said images without needing to use an energy-filter with sub-volt resolution. 20. The apparatus of claim 19, further comprising a data processing system configured to calculate a measure of variation of an imaged aspect of a feature with respect to the potential difference. 21. The apparatus of claim 20, wherein the imaged aspect of the feature comprises an intensity of the feature. 22. The apparatus of claim 20, wherein the imaged aspect of the feature comprises an apparent size of the feature. 23. The apparatus of claim 20, wherein the data processing system is further configured to determine a process-related parameter based at least in part on the calculated measure of variation of the imaged aspect of the feature with respect to the potential difference. 24. The apparatus of claim 23, wherein the control system is further configured to obtain image data from multiple locations on the substrate. 25. The apparatus of claim 24, wherein the data processing system is further configured to build a map showing variation of the process-related parameter over the substrate. 26. The apparatus of claim 19, wherein the control system varies the potential difference by changing a voltage applied to a stage holding the substrate. 27. The apparatus of claim 19, wherein the control system varies the potential difference by changing a voltage at the source of the incident electron beam. 28. The apparatus of claim 19, wherein the apparatus generates a second incident electron beam for substrate charge control. 29. The apparatus of claim 19, further comprising a UV light source configured to illuminate at least a portion of the substrate for surface charge control. 30. A method of inspecting a substrate using electrons, the method comprising:obtaining multiple images of a region of the substrate using mirror-mode electron-beam imaging at a range of voltage differences between an electron source and a substrate, each said image corresponding to one said voltage difference;storing mirror-mode image data corresponding to the multiple voltage differences;scattered electron imaging of the region;storing the scattered electron image data; andcomparing gray levels of a feature in the region in the mirror-mode image data to the scattered electron image data of the feature,wherein the method provides sub-volt voltage contrast between said images without needing to use an energy-filter with sub-volt resolution. 31. The method of claim 30, wherein the scattered electron imaging comprises secondary electron imaging. 32. The method of claim 30, wherein the scattered electron imaging comprises backscattered electron imaging. |
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claims | 1. An electron microscope, comprising:an electron gun;an objective lens configured to focus an electron beam emitted from the electron gun on a sample;a spherical aberration corrector configured to correct a spherical aberration of the objective lens; anda phase plate configured to have a thickness which continuously changes in a radial direction,wherein a phase difference due to a difference in electron beam path of the electron beam passing through the objective lens is eliminated by the phase plate. 2. The electron microscope according to claim 1, further comprising a mechanism which is capable of automatically performing axis adjustment of a phase adjusting mechanism. 3. The electron microscope according to claim 1 wherein the phase plate is a conductive crystalline phase plate. 4. The electron microscope according to claim 1, wherein the phase plate is a conductive amorphous phase plate. 5. The electron microscope according to claim 1, further comprising a phase plate supporting mechanism including:a mechanism for adjusting a tilt, a position, and/or a height of the phase plate; anda mechanism for adjusting temperature. 6. The electron microscope according to claim 1, further comprising—a phase plate chamber,an electron beam passage port,a gas introduction pipe,a gas flow adjustment valve, andan exhaust apparatus, for adjusting a gas species pressure in the vicinity of the phase plate. 7. An electron beam phase adjuster, which is used for the electron microscope according to claim 1. 8. The electron microscope according to claim 1, wherein the electron microscope is any one of: a phase difference electron microscope; a transmission electron microscope which is configured by combining the phase difference electron microscope and a spherical aberration corrector; a scanning electron microscope; a scanning transmission electron microscope; and a transmission electron microscope. 9. The electron microscope according to claim 1, wherein the phase plate has a potential which changes in a radial direction. 10. The electron microscope according to claim 1, further comprising:a phase plate supporting mechanism configured to have a function of adjusting a potential. 11. The electron microscope according to claim 1, further comprising:a mechanism for performing, if a change in phase does not uniformly occur on the same concentric circle, correction of the change in phase. |
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description | The United States Government has rights in this invention pursuant to Contract No. DE-AC02-06CH11357 between the U.S. Department of Energy and the University of Chicago representing Argonne National Laboratory, and funding from the U.S. Environmental Protection Agency. One or more embodiments of the present invention relates to methods and systems for the mitigation of radionuclide contaminants from contaminated surfaces. The method is particularly useful for mitigation with the intended purpose of quickly restoring critical infrastructure and operational activities following a radionuclide dispersion event. During a radionuclide dispersion event, a radionuclide dispersion event being a release of radioactive contaminants through an act of malice, nuclear reactor accident, or otherwise, large swaths of area and equipment may be contaminated. An exemplary act of malice is the activation of a radiological dispersal devise, commonly known as a “dirty bomb,” where an amount of radioactive material is used to maliciously contaminate people, equipment, and/or the environment without a nuclear explosion. “‘Dirty Bombs’: Technical Background, Attack Prevention and Response, Issues for Congress” J. Medalia, 7-5700, Jun. 24, 2011, Congressional Research Service.” In the course of such a radiological dispersion event, radionuclide contaminants may be spread across large areas. Surfaces of critical infrastructure and equipment such as fire houses, medical facilities, and emergency response tools and equipment onto which the radionuclide contaminants rest may become contaminated, in turn compromising response efforts by emergency response officials. In addition, some public services such as drinking and wastewater treatment, electrical power distribution, etc. may be disrupted. In such an event, it is important to deploy mitigation efforts to reduce levels of radioactive contaminants by removal or decontaminating certain areas in order to restore response activities and public services. Disclosed herein is a method and system useful for the mitigation of radionuclide contamination from surfaces bearing radionuclide contaminants. The method and system are particularly useful to rapidly return components such as fire-fighting equipment to service following a radionuclide dispersion event while also avoiding further radionuclide contaminant spread beyond the original deposition area and minimizing the amount of additional materials contaminated during the mitigation processes. The method and system are rapidly deployable, cost-effective measures to mitigate critical infrastructure and equipment for the purpose of restoring that infrastructure and equipment to operational activities after a radiological release. Application of the radionuclide contaminant mitigation method or utilization of the system has the potential to reduce the level of radionuclide contaminants on surfaces. They may be performed on both a small scale as on tools, detectors, and personal protective equipment, as well as on a large scale as would be the case with critical infrastructure or equipment. Thus, by performing the method or utilizing the system, the exposure to radiation by emergency workers, responders, and the general population who are in close proximity to those surfaces is lessened. The method comprises applying a carrier solution comprising a cation onto a surface bearing a radionuclide contaminant, such that the radionuclide contaminant enters the solution to form a laden solution. The method then employs contacting the radionuclide contaminant laden solution with solid sequestering agents that bind to and immobilize at least a portion of the radionuclide contaminant. In providing another binding material, the mobility of the radionuclide contaminant through the environment is lessened as a whole. Mitigation with respect to the method refers to the removal of at least a portion of the radionuclide contaminant from a bearing surface. Mitigation typically occurs shortly after a radiological dispersion event. This removal is necessary in order to restore critical infrastructure such as fire houses to a level that the infrastructure can be utilized. Note that mitigation differs from decontamination; a long term activity designed to clean-up the contaminated infrastructure to acceptable near background levels. Thus, these mitigation methodologies may not be as effective in decreasing radionuclide contaminant radiation levels to background level as those methodologies used for final decontamination. Nevertheless, during mitigation, speed and economy at which methodologies or systems can be deployed and completed may be of equal importance relative to effectiveness, and may also impact the effectiveness of follow-on decontamination for longer term recovery. Radionuclide contaminants include the isotopes of stable elements that produce alpha, beta, or gamma radiation through their radioactive decay. Radioisotopes are rarely available in quantities sufficient to cause harm outside of highly restricted areas such as nuclear reactors. Thus, while there are approximately 3,715 different identified radionuclides, there is a relatively small number likely to be spread in a radiological dispersion event based on quantities available, key physical and chemical characteristics such as half-life, and decay activity. Those radioisotopes of greatest concern are listed in Table A below. Of those listed in Table A, the radioisotopes available most likely to be in civilian control and thus most likely to be spread by a radiological dispersion event include Co-60, Cs-137, Ir-192, Sr-90 and Am-241. TABLE AMode ofRadioisotopeDecayAm-241αCd-109x-rayCf-252αCo-60β-γCs-137β-γCs-134β-γFe-55x-rayGd-153x-rayHo-166β-γI-125x-ray-γI-131β-γIr-192β-γKr-85β-γLu-177β-γNi-63βP-32βP-33βPd-109β-γS-35βSe-75γSr-90βW-188β Radionuclide contaminants may be in a variety of physical forms such as powders, clad with ceramics (reactor fuel rods), etc. The most likely forms are cesium chloride, cesium oxides, cesium alumino-silicates, other cesium ceramic materials, strontium fluoride, strontium oxides, strontium titanates, cobalt metal or metal alloy, iridium metal or metal alloy, and mixed fission product and actinide nitrates. Of these, the alumino-silicates, ceramics, titanates, metals and metal alloys are insoluble in water and would likely persist as particulate material after dispersion into the environment. As such, decontamination would involve the physical removal of particles from contaminated surfaces rather than chemical desorption of radioactive contaminants from surface chemical sorption sites. The cesium chloride and oxides, and the strontium fluorides and oxides are soluble in aqueous solutions. After dispersion during a radionuclide dispersion event, these radionuclide contaminants can dissolve in the water present within a building material (e.g., pore water), from a precipitation event, or from contact with bulk water (e.g., fire hose, ocean spray). This dissolution of the radionuclide contaminants allows them to easily become chemically bonded onto the surface material, primarily through ion exchange reactions with the surface. In terms of chemical form, those radionuclides that persist as small particulate are easier to mitigate than those that have dissolved in water and reacted chemically with the surface. As illustrated in FIG. 1, the chemical and physical characteristics of the surface on which the radionuclide contaminant rests has a significant impact on the percentage of radionuclide contaminant that can be mobilized by the carrier solution. Asphalt, brick, limestone, granite, and concrete are representative of porous urban construction materials present in critical infrastructure such as roadways, hospitals, and public works facilities. Different surfaces frequently lead to different sorption mechanisms and dissimilar surfaces promote mobilization of the radionuclide contaminants while using the method or system. The mitigation of radionuclide contamination as performed by the method and system, when put into the context of the overall response to a radiological dispersion event, makes a significant difference in the outcome in several areas. The resulting reduction allows emergency responders to conduct operations for longer periods of time by reducing their cumulative dose of radiation while working in a “hot” zone. This is critical in the early phases of a radiological dispersion event when the number of first responders will be limited. In an exemplary radiological dispersion event where there would be a release of 1,000 Curies of Cesium-137 (50 grams) over an urban area of 2.10 km2, a population of 38,000 individuals if not removed from the area would experience an increase of 233 cases of cancer including 159 fatal cases, from a first year of exposure alone. “‘Dirty Bombs’: Technical Background, Attack Prevention and Response, Issues for Congress” J. Medalia, 7-5700, Jun. 24, 2011, Congressional Research Service.” During emergency response operations, a 50% reduction of a radionuclide contaminant such as Cs-137 from the surface of a piece of equipment and the resulting reduction in radiation exposure will allow emergency response workers to remain in the area twice as long performing twice as much life-saving activity at a time when qualified workers represent a scarce resource. The lower exposure also benefits workers involved in restoring the area to its original condition during late-phase recovery activities by removing the easily mobilized radionuclide contaminants from the area before more aggressive decontamination procedures are implemented. Further, reducing the contamination level even 30% at the beginning of a radiological dispersion event response may result in significant savings later, because the longer the radionuclide contaminant remains in contact with some common materials, the more aggressive the approach needed to decontaminate them—for example grinding off the surface may be necessary, which can be logistically more challenging and destructive and further results in a higher cumulative radiation dose exposure to the work force. Finally, reduction in radionuclide contaminants via this method correlate to potentially significant reductions in the volumes of radiologically impacted wastewater. Namely, the method immobilizes and allows the removal of radionuclide contaminants preventing their spread through a watershed. The method disclosed provides radionuclide contamination mitigation by applying an aqueous carrier solution comprising a cation to a surface bearing a radionuclide contaminant, causing the radionuclide contaminant to enter solution to form a laden solution, then contacting the laden solution with a solid sequestering agent to bind to the radionuclide contaminant to form a laden sequestering agent. The removal and sequestration of the radionuclide contaminant from the contaminated surface leads directly to a reduction in the amount of radiologically-impacted equipment, critical infrastructure, and the environment. The method is able to be performed economically with materials quickly available in the event of a radiological dispersion event. In a second embodiment, the method described provides radionuclide contamination mitigation by forming a brine; mixing a carrier solution comprising the brine and water; applying the carrier solution to a surface bearing a radionuclide contaminant causing a laden solution to be formed, and contacting the laden solution with a sequestering agent causing the formation of a laden slurry comprising laden sequestering agent and reformed carrier solution. In another embodiment, a system provides for the mitigation of a surface bearing a radionuclide contaminant. The system provides an applicator for applying a carrier solution to a surface bearing a radionuclide contaminant and a container for a sequestering agent; The following description is provided to enable any person skilled in the art to use the invention and sets forth the best mode contemplated by the inventor for carrying out the invention. Various modifications, however, will remain readily apparent to those skilled in the art, since the principles of the present invention are defined herein specifically to provide a method and system for the mitigation of radionuclide contamination of a surface. Embodiments of the invention are directed to a method and system for mitigation of a radionuclide contaminant from a surface. The method for mitigation generally comprises applying a carrier solution comprising a cation and water, to a surface bearing a radionuclide contaminant, resulting in mobilization of the radionuclide contaminant into the carrier solution to form a laden solution, and contacting the laden solution with a solid sequestering agent such that the radionuclide contaminant in the laden solution binds to the sequestering agent to form a laden sequestering agent and generating a laden slurry comprising the laden sequestering agent and reformed carrier solution. Further embodiments include forming a brine solution comprising the cation and water. Still further embodiments include retaining the sequestering agent in a permeable container. Another embodiment includes performing a separating operation on the laden slurry to separate the laden sequestering agent from the reformed carrier solution. Still further embodiments include recycling the carrier solution to again apply it to a radionuclide contaminant bearing surface. One embodiment of the invention is a system for the mitigation of a radionuclide contaminant from a surface, the system comprising: an applicator for applying a carrier solution comprising a cation and water, and a container for a sequestering agent. Radionuclide Contaminant: As noted above, radionuclide contaminants are spread across a surface during a radionuclide dispersion event. Radionuclide contaminants of greater concern are those identified in Table A. The method and system, in particular, are useful for the mitigation of Cs-137 and Sr-90. Surface: The surfaces bearing a radionuclide contaminant vary to include non-porous, metallic surfaces such as hand tools, porous surfaces such as concrete. Further, the surfaces may be composed of inorganic or organic materials such as in textiles. The chemical and physical characteristics of the surface on which the radionuclide contaminant rests has a significant impact on the percentage of radionuclide contaminant that can be mobilized by the carrier solution. Depending on the character of the contaminated surface, the amount of radiation reduction via the method or system may be as high as 90-100% for non-porous surfaces. For porous surfaces where the radionuclide contaminant penetrates or binds with high affinity, the amount of radiation reduction may still reach 30-50%. (FIGS. 2 and 3) Solvent: The carrier solution generally comprises water and a cation. Aqueous solutions are beneficial as they are quickly accessible in large quantities through supplies such as municipal supply water from water hydrants. Where municipal supplies are not available, water may be sourced from stores such as reservoirs, cisterns, and tanks. In addition to their benefits due to availability, the aqueous carrier solution is beneficial as the solvent itself does not further contaminate the wash (mitigation) area. Cations: In addition to the aqueous solvent, the carrier solution is further comprised of cations in the aqueous solvent. The cations are thought to promote ion exchange replacement reactions with radionuclide contaminants bound to the contaminated surfaces. Cations are preferentially selected for use in the method and system based on charge density and hydrated radii. Further, such cation characteristics may be used as a basis for selection through relative comparison to the radionuclide contaminants to be mitigated. For example, the cation K+ may be selected for the mitigation of cesium radionuclides, based on the similarity of charge density and ionic radius. Preferably, the ionic radius is within about 20% difference of the cation selected and the radionuclide to be mitigated. Preferred cations for use in the method and system include potassium (K+), ammonium (NH4+), and sodium (Na+). Other suitable cations include barium (Ba2+), calcium (Ca2+), and magnesium (Mg2+). Fortunately, these cations are available worldwide in bulk quantities as potassium chloride, potassium nitrate, ammonium nitrate, and ammonium chloride as well as other forms. Further, mixtures of cations are contemplated. For example, carrier solutions may include both K+ and NH4+ in solution via the soluble salts KCl and NH4Cl. A salient aspect of the invention is the concentration of the cation in the carrier solution. If the concentration is too low, the carrier solution fails to promote the radionuclide contaminant entering solution via the ion-exchange route (FIG. 4). If the concentration is too high, the method and system risk the carrier solution causing premature collapse of the sequestering agents as the cations compete with the radionuclide contaminant. Additionally, excessively high concentrations increase the economic and logistical burden, causing unnecessarily large stockpiling of supplies for increasingly diminishing returns of effectiveness. As used in the method, the concentration of the cations is preferentially in a range from about 0.1 M to about 1.0 M. More preferentially, the range is from about 0.1 M to about 0.5 M. Most preferentially, the range is from about 0.25 M to about 0.5 M. Surfactant: In one embodiment, the carrier solution further comprises a surfactant to facilitate mobilization of a radionuclide contaminant into the carrier solution. Some surfaces promote greater bonding of the radionuclide contaminant to the surface. For example, radionuclide contaminants are able to form stronger bonds to certain aggregates in concrete. With respect to other surfaces, a hydrophobic character discourages sufficient contact with the aqueous carrier solution to allow the radionuclide contaminant to enter the carrier solution for a laden solution. In these situations, addition of a surfactant to the carrier solution acts to improve the wettability of hydrophobic surfaces such as asphalt to promote ion exchange. A wetting liquid forms a contact angle with the solid that is less than 90°, whereas a nonwetting liquid creates a contact angle between 90° and 180° with the solid (FIG. 5). Surfactants found suitable as additives in the carrier solution are anionic and neutral surfactants since these are widely available and are compatible with use outdoors. For instance, cationic surfactants can be disruptive to cell membranes. Exemplary anionic surfactants available worldwide and in large quantities to use in the method include sodium stearate, sodium dodecyl sulfate, ammonium dodecyl sulfate, cetyltrimethyl ammonium bromide, sodium dodecylbenzenesulfonate and other linear alkylbenzene sulfonates (LAS). Exemplary non-ionic surfactants include TRITON X-100 [polyethylene glycol p-(1,1,3,3-tatramethylbutyl)-phenyl ether] and other polyoxyethylene glycol octylphenol ethers, polyoxyethylene glycol alkyl ethers and polyoxypropylene glycol alkyl ethers; polysorbate 80 (aka TWEEN 80) or other polyoxyethylene glycol sorbitan alkyl esters and sorbitan alkyl esters (e.g., Spans); MEGA 10 (N-decanoyl-N-methylglucamine); and glucoside alkyl ethers, specialty surfactant products such as the Aqueous Film-Forming Foam Concentrates (AFFF) used by firefighters to combat chemical fires combine fluoro- and hydrocarbon-surfactants in a proprietary manner and may be appropriate to use in this method. More preferable surfactants are sodium dodecyl sulfate and ammonium dodecyl sulfate. As used in the method, the concentration of the surfactants is preferentially in a range that stays below the critical micelle concentration under solution conditions to avoid formation of suds, emulsions, or precipitates or from about 10−6 M to about 5×10−2 M. More preferentially, the range is from about 1×10−5M to about 5×10−2 M. Most preferentially, the range is from about 0.5×10−4M to 5×10−3 M. Other additives may be advantageous for addition to the carrier solution. Additives such as foaming and gelling agents may be added. When in the carrier solution, they may limit the spread of the carrier solution after application onto the surface bearing a radionuclide contaminant. Water soluble dyes may be added to the carrier solution as a visual indicator to track the spread of the carrier solution after application onto the contaminated surface. Brine: In one embodiment, a brine is used to form the carrier solution. A brine is a more concentrated solution of water, the cation, and any additives of a respective carrier solution. To ultimately form the carrier solution for use in the method, the brine is mixed with water from a bulk source to dilute the brine to reach the desired carrier solution concentration. The carrier solution is then available to be applied to the contaminated surface. Formation of the carrier solution through the brine embodiment may be advantageous in that it is possible to form the brine in smaller tanks where time, space, or distance is a concern. The brine may then be mixed with the bulk water supply in the immediate mitigation area through means such as an eductor system commonly available to fire departments. These eductors draw a secondary fluid into the main water line without the need for specialized pumps. Additionally, eductors commonly have metering valves making it possible to regulate the percentage of brine taken up as the secondary fluid into the main water line. In one illustration, emergency response workers can quickly prepare a 15 gallon saturated brine of KCl (342 g/L at 20° C.) in a vessel. An inline eductor connected to a fire hose and with the secondary takeup tube in the brine vessel is set to draw a 6% mixture of brine/water at a 250 gallon/minute flow rate delivered through the nozzle. The worker then engages the eductor, applying the resulting approximate 0.25 M carrier solution onto a contaminated surface. In this manner, the emergency workers can apply 250 gallons of carrier solution onto a surface in one minute, without the need for extra equipment since solution tanks and eductors are often already available locally. Applying: The method involves a step of applying a carrier solution to a surface bearing a radionuclide contaminant. During applying, the carrier solution is put into physical contact with the contaminated surface, such that the radionuclide contaminant is mobilized by entering into the carrier solution, to form a laden solution comprising the carrier solution and the radionuclide contaminant. With respect to Cs-137 and Sr-90, the radionuclide contaminants could be mobilized through dissolution into the carrier solution. Application of the carrier solution can be accomplished through such exemplary actions as spraying the carrier solution onto the contaminated surface, dumping the carrier solution from a bucket or sprayer as used in forest fires, and flowing the carrier solution across a horizontal surface such as a parking lot, and applicators capable for performing these actions (hoses, sprayers, fluid dumps, etc.). Further, relatively small scale items such as tools may be immersed in a carrier solution tank. It is advantageous that the application be through a means not so forceful as to mobilize the radionuclide contaminant back into the atmosphere or otherwise away from the carrier solution, but instead promote the radionuclide contaminant to enter solution, forming the laden solution. Further, it may be advantageous to perform application in a manner that prolongs exposure of the radionuclide contaminant bearing on the surface to the carrier solution, to encourage complete entry into solution where the percent of entry increases with time (FIG. 6). Laden Solution: Once radionuclide contaminant has been mobilized from the contaminated surface into the carrier solution, a laden solution is formed. The laden solution comprises at least a portion of the radionuclide contaminant and the carrier solution. Sequestering Agents: Sequestering agents are solids capable of binding to a radionuclide contaminant or form precipitates when combined with the radionuclide or after adjustment of the solution conditions (e.g., pH, temperature, ionic strength). Binding may be chemical, physical, or both. The sequestering agent may be in the form of particulate solids, membranes, or a combination of both. It is recognized by those skilled in the art to improve the efficiency of separations by employing successive batch separations at lower slurry concentration (e.g., at 40 mg/mL or less) to reduce the amount of total sequestration material and to remove exhausted sequestration material in one batch unit at a time. An alternative method is to employ a bed of material where the radioactive laden solution permeates up or down the bed before being transferred to another bed unit. Possible sequestering agents include clays such as montmorillonite (a primary constituent of bentonite and Fuller's earth), vermiculite, illite, Halloysite, sepiolite, kaolinite, and palygorskite (a primary constituent of Fuller's Earth); natural (e.g., chabazite) and synthetic (e.g., crystalline silico-titanate and its derivatives) zeolite; soils, rocks; industry byproducts such as coal and municipal waste fly ash, building and roadway demolition materials; activated carbon or other natural and synthetic organic and inorganic ion exchange materials. Sequestering agents that form precipitates include the cyanoferrates (e.g., Prussian blue) for cesium and carbonate rich minerals such as limestone and hydrotalcite. Membranes include reverse osmosis. Combination sequestering agents include the resin wafer electrodialysis. Preferred sequestering agents include clays. Clays are stable over time and are available in large quantities throughout the world. Clay minerals are hydrous aluminum silicates arranged in the form of layered sheets (phyllosilicates) with variable amounts of cations such as iron, magnesium, alkali metals, alkaline earths. There are several groups of clays: kaolinite, montmorillonite-smectite, illite, and chlorites, although chlorites are often categorized as a separate type of phyllosilicate material. Clay minerals are further classified as 1:1 or 2:1 to describe the types of tetrahedral silicate sheets and octahedral hydroxide sheets they are composed. A 1:1 clay would consist of one tetrahedral sheet and one octahedral sheet like those of the kaolinite clays. A 2:1 clay consists of an octahedral sheet between two tetrahedral sheets as in the vermiculites and montmorillonites. Depending on the composition of the tetrahedral and octahedral sheets, the layer will have either a neutral charge or a net negative charge. A net negative charge is caused by replacement of higher oxidation state cations with lower oxidation states during clay formation. The excess charge is balanced by the sorption of interlayer cations, commonly the alkali metals and alkali earth metals. The interlayers will also contain water that leads to the common property of swelling in clays. Clays have advantageous characteristics as sequestering agents in that they have high swelling ability, low hydraulic conductivity, and high cationic sorption capacity. Additionally, clays are ideal as sequestering agents because cations with low hydration energy, for example radionuclide cesium contaminants, undergo dehydration in the interlayer of the clays and promote layer collapse, thus radionuclide contaminants become mechanically fixed within the clay's interlayers. Preferential clays exhibit high sorption of likely radionuclide contaminants, with performance as illustrated in FIG. 7. Preferential clays are montmorillonite and vermiculite because of their availability and affinity for radioactive cations. Most preferential is vermiculite. Contacting the Laden Solution with Sequestering Agent: In contacting, the laden solution comprising radionuclide contaminant in the aqueous carrier solution is physically contacted with the sequestering agent, such that at least a portion of the radionuclide contaminant may bind to the sequestering agent. When at least a portion of the radionuclide contaminant from the contaminated surface binds to the sequestering agent, a laden sequestering agent is formed. Further, a laden slurry is formed comprising the laden sequestering agent and reformed carrier solution. The reformed carrier solution is defined as a solution comprising formerly laden solution where at least a portion of the radionuclide contaminant has left solution through binding to the sequestering agent. Thus, by binding the radionuclide contaminant to the sequestering agent, the radionuclide contaminant is removed from solution and the reformed carrier solution is available for repeated use. Preferably, at least 90% of the radionuclide contaminant in the laden solution binds to the sequestering agent in the solution transition from laden slurry to reformed carrier solution In function, the sequestering agent, through binding, removes the radionuclide contaminant from solution and enables the radionuclide contaminants easy disposal. As a result, the once mobile radionuclide contaminant sitting on a surface has been immobilized and prevented from being further spread through the area. There are many approaches for contacting the laden solution with the sequestering agent. Where large applications of the method are being performed, or the system being utilized, as would be required for example in the mitigation of heavy equipment or building infrastructure, the sequestering agent may be spread across the ground downgrade from the application area. This may be performed by dumping the sequestering agent from a vehicle such as a truck and spreading the sequestering agent using shovels. The intent is to have the solutions run down from the infrastructure to come in immediate contact with the sequestering agent so that the radionuclide contaminants are quickly bound and immobilized. Any solution coming into contact with the sequestering agent would be stripped of at least a portion of the radionuclide contaminant by the sequestering agent, forming the laden sequestering agent and generating a laden slurry. In another approach to contacting laden solution with sequestering agent, the sequestering agent may be premixed into the carrier solution. When premixed, the resulting viscous slurry/mud-type mix may be ideal for flowing across highly graded or pitched surfaces, to both increase the duration the carrier solution is in contact with the contaminated surface as well as increase controllability of the fluids during mitigation operations. In yet another approach, the sequestering agents may be used to fill a container, where a portion of the container is made of a material permeable to the carrier solution, and in turn the reformed carrier solution. Further, the containers may also be permeable to the laden solution, and in turn the reformed carrier solution. In an exemplary use of such a container, the containers may be used to form a system of berms to create reservoirs. The containers would be able to substantially contain laden solution, while any laden solution leaching through the barrier would be stripped of at least a portion of the radionuclide contaminant, essentially forming reformed carrier solution. This approach may be exemplary performed by formation of a reservoir made by filling collapsible construction containers or gabions commonly stockpiled for construction, military, and flood applications. Gabions are preferentially filled with the sequestering agent; however, availability may necessitate at least partial fill such as rocks, dirt, or sand. After formation of the reservoir, the carrier solution is applied to the contaminated surface and the resulting laden solution is contained within the reservoir. As time and materials allow, additional sequestering agent may be added to the reservoir to bind radionuclide contaminants. Additionally, containers of sequestering agent may be placed atop catch basins, sewer caps, and utility manhole covers to prevent extensive intrusion of contaminated waters into the underground tunnels. In the above reservoir approach, containment of the laden solution within the reservoir vessel also allows additional sequestering agent to be mixed with the laden solution either inside or outside of the reservoir. When mixing occurs within the reservoir, the resulting laden slurry can then be pumped from the reservoir or allowed to air dry in a controlled manner as the radionuclide contaminants are bound to the sequestering agents. The dried laden sequestering agent could be disposed of as time, facilities, and personnel allow. If the laden solution is to be removed from the reservoir prior to mixing, the mixing operation may occur as part of the removal step (mixing while pumping) or later. When contacting the sequestering agents with the laden solution, it is preferable to limit the volume of laden solution coming into contact with the sequestering agents. As the sequestering agents become increasingly loaded with contaminant, the ability to absorb radionuclide contaminants is diminished. Exemplary means to monitor the volume of laden solution coming into contact with the sorbent solution include employing holding tanks or storage reservoirs and mixing or metering the laden solution sequentially through the sequestering agent or to permit the laden solution to penetrate a bed of sequestering agent and then transferring the solution to another bed for further treatment. It may be preferred to limit the volume of laden solution contacting to a ratio of 1 mL/0.01 g or 1 L/0.01 kg or 100 L/kg in successive batch reactions. A more preferable range is to no more than 10 L/kg. Retaining In order to perform certain operations, a retaining step is required to retain the mobilized radionuclide contaminant after it has left its respective surface. Retaining means physically-confining or keeping control of the associated material, be it laden solution, laden sequestering agent, regenerated carrier solution, etc. As noted above, retaining can be accomplished by forming a system of berms to create a reservoir. In another example, retaining is accomplished through performing the method or confining the system to a sealed container or containers such as a series of washbasins. Separating In another embodiment, a separating step is performed. In separating, after contacting the laden solution to the sequestering agent such that at least a portion of the radionuclide contaminant binds to the sequestering agent to form the laden sequestering agent, a separating operation is performed to separate the regenerated carrier solution from the laden sequestering agent. In effect, separating substantially divides the laden sequestering agent from the fluids. In instances where there is sufficient time, separating may be performed by allowing the sequestering agent to settle to the bottom of the vessel. In that process, the regenerated carrier solution may be skimmed from the top to be recycled or disposed in another manner. Preferentially, separating is accomplished by carrying out filtration. Filtration is commonly known as a mechanical or physical operation to separate solids from liquids. In one variation of filtration, a centrifugal filter is utilized. Centrifugal filters have no moving parts excluding the pump that feeds the separator. Centrifugal action forces solid material into a purge zone, separating solid materials from clarified fluid product. Centrifugal filtration units capable of operating at very high throughput (millions of gallons per day) on high slurry concentrations are commercially available, easily sourced, and are modularized for movement and operation on vehicles. Filtration may also be performed by mechanical filtration as with simple bag filters. Recycling Additionally, separation operations may be conducted in series. For example, the clarified fluid from a centrifugal filter or series of centrifugal filters may be separately fed to another unit such as a bag filter for secondary or final clarification. As a result of the described separation operation, a substantial portion of the laden sequestering agent is retained by the filters while the clarified solution, which is substantially regenerated carrier solution with only relatively small residual amount of radionuclide contaminant, is separated. After separating, recycling operations may be performed. In recycling, the carrier solutions regenerated during a separating step are used during a subsequent application step. By performing recycling, the total volume of contaminated fluids is reduced. The reduction of fluid used for solution and generated during practicing the method relieves the stress on the water supply and reduces the total amount of fluid waste requiring disposal. In an exemplary combination of the above embodiments, a carrier solution comprising water and a cation is formed using a brine. The brine is combined with bulk water through an eductor, which also serves for applying the resulting carrier solution to a surface contaminated with radionuclide contaminants. The laden solution formed as the radionuclide contaminants enter the carrier solution is directed into a reservoir formed from gabions filled with sequestering agents. When time allows, the pool of laden carrier solution is pumped from the reservoir and mixed with additional sequestering agents, contacting the laden solution with sequestering agents forming the laden slurry comprising laden sequestering agents and regenerated carrier solution. The laden slurry is then separated first by centrifugal filtration to remove the bulk of the laden sequestering agents. The filtrate is then filtered again, this time through a bag or membrane filter, to remove the remainder of the sequestering agents including the laden sequestering agents. The solids are then disposed. The regenerated carrier solution resulting from the separation process is then recycled back to mitigation area for a repeated application to the contaminated surface. The removal and sequestration of the radionuclide contaminant from the contaminated surface leads directly to a reduction in the amount of radiologically-impacted equipment, critical infrastructure, and the environment. The method is able to be performed economically with materials quickly available in the event of a radiological dispersion event. The disclosure further contemplates a system for the mitigation of a system to provide the mitigation of a radionuclide contaminant from a surface. The system comprises at least an aqueous carrier solution comprising a cation, and a sequestering agent. The system may further comprise individually or in combination: at least one permeable container for retaining sequestering agent; a separator, such as a centrifugal filter or membrane filter; an applicator, such as a sprayer, fire house, or eductor system. Having described the basic concept of the invention, it will be apparent to those skilled in the art that the foregoing detailed disclosure is intended to be presented by way of example only, and is not limiting. Various alterations, improvements, and modifications are intended to be suggested and are within the scope and spirit of the present invention. Additionally, the recited order of the elements or sequences, or the use of numbers, letters or other designations therefore, is not intended to limit the claimed processes to any order except as may be specified in the claims. All ranges disclosed herein also encompass any and all possible sub-ranges and combinations of sub-ranges thereof. Any listed range can be easily recognized as sufficiently describing and enabling the same range being broken down into at least equal halves, thirds, quarters, fifths, tenths, etc. As a non-limiting example, each range discussed herein can be readily broken down into a lower third, middle third and upper third, etc. As will also be understood by one skilled in the art all language such as “up to,” “at least,” “greater than,” “less than,” and the like refer to ranges which can be subsequently broken down into sub-ranges as discussed above. Accordingly, the invention is limited only by the following claims and equivalents thereto. All publications and patent documents cited in this application are incorporated by reference in their entirety for all purposes to the same extent as if each individual publication or patent document were so individually denoted. |
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048225584 | description | DESCRIPTIN OF THE PREFERRED EMBODIMENTS Turning to FIG. 1, a pressurized water reactor (not shown in detail) has a hemispherical bottom end wall 1 and supports a horizontal lower core plate 2, on the top face 3 of which there are supported fuel assemblies, of which one is shown and generally designated at 4. The fuel assembly 4 comprises a plurality of fuel rods 5 supported at their lower end by a bottom nozzle generally designated at 6. The nozzle 6 comprises a horizontal platform 7 having a bottom face or underside 8, as well as spaced legs 9 which extend downwardly from the platform 7 and stand on the top face 3 of the lower core plate 2, thus supporting upright the entire fuel assembly 4. Through axially aligned vertical passages 10, 11 and 12 provided, respectively, in the reactor wall 1, the lower core plate 2 and the platform 7 there passes a thimble 13 which is a slender stainless tube reaching to the top part of the fuel assembly 4 at its upper end and projectin out of the reactor wall 1 in a downward direction to be associated with measuring instrumentation, not shown. The thimble 13 serves the purpose of guiding therein a probe (also not shown) which travels vertically along the length of the fuel assembly 4 to sense the neutron flux. The thimble 13 is axially moved out of the fuel assembly during refueling and is subsequently reinserted into the fresh fuel assembly. In the zone between the reactor wall 1 and the lower core plate 2 the thimble 13 is surrounded by a permanently installed sleeve construction 14 whose securing elements and thimble guide components are generally designated at 14a. A thimble guide 15 constituted by an elongated sleeve-like component is inserted at the upper end of the passage 11 of the lower core plate 2 and extends vertically upwardly, in alignment with the passage 12 provided in the nozzle platform 7. The thimble guide 15 is rigidly held--for example, by a tightened threaded connection--in the core plate 2. The thimble guide 15 serves as a sliding guide and for positioning and stabilizing the thimble 13. Above the nozzle platform 7 the thimble 13 is positioned within an instrumentation tube 16 extending upwardly from the platform 7 and forming a permanent part of the fuel assembly 4. On the thimble guide 15 there is inserted an only symbolically shwon thimble guide extender generally designated at 17 for the particular purpose of providing a shield about the thimble 13 in the zone situated between the top of the thimble guide 15 and the underface 8 of the platform 7. The thimble guide extender 17 constitutes a first preferred embodiment of the invention, now to be described in detail with reference to FIG. 2 in which the extender 17 is shown in its installed state. The extender 17 has a longitudinal axis A and comprises a lower fitting 18 having a flange 19 and an upper fitting generally designated at 20, being vertically spaced from and generally in alignment with the lower fitting 18. The upper fitting 20 has a hollow cap 21 having an inner radial terminal shoulder 21a and provided with a passage 22 adapted to closely surround the thimble 13 (not shown in FIG. 2) and an integral or welded sleeve 23 having a radially inwardly extending flange 24 of downwardly flaring configuration at 25 and a narrow, annular contact area 25a. A cylindrical bellows 26 of circumferentially closed configuration is sealingly welded at its top to a radial lower face of the sleeve 23 and is, at its bottom, sealingly welded to a radial top face of the flange 19 of the lower fitting 18. The bellows 26 is resiliently deformable in several directions and may be made, for example, of No. 316 stainless steel. Within the bellows 26, coaxially therewith, there is arranged a retainer ring generally designated at 27 having an upper terminal radial face 31a. Approximately the lower two-thirds of the length of the retainer ring 27 is slit to form a plurality of circumferentially arranged, radially resilient legs 28 each having an inwardly extending bottom projection 29. The retainer ring 27 further has at its bottom, approximately at the axial location of the leg projections 29, radially outwardly extending ring segments 30. At the top of the retainer ring 27 there is formed a radially outwardly extending flange 31. The retainer ring 27 is, at its lower part, in a stabilizing engagment with the inner face of the lower fitting 18 by means of the ring segments 30. At its upper portion, the retainer ring 27 is in circumferential engagement with the surrounding sleeve 23 of the upper fitting 20. The circumferential contact between the sleeve 23 and the retainer ring 27 is along the narrow, annular contact area 25a. Such small area contact permits a rocking motion of the upper fitting 20 relative to the retainer ring 27 about an axis transverse to the longitudinal extender axis A for purposes to become apparent as the specification progresses. A cooperation between a lower radial face of the flange 31 of the retainer ring 27 and an upper radial face of the flange 24 of the sleeve 23 forming part of the upper fitting 20 limits an outwardly directed axial motion of the upper fitting 20 on the retainer ring 27 as urged by the bellows 26. The cylindrical bellows 26 is spacedly and fully surrounded by a cylindrical shield 32 which is secured at its bottom part to the lower fitting 18 and which has an open top through which projects the sleeve 23 of the upper fitting 20. The thimble guide extender 17 is installed on the thimble guide 15 previously secured to the lower core plate 2 from above, in the absence of the fuel assembly 4. For this purpose, the extender 17 is guided downwardly such that the tapered top part of the thimble guide 15 is introduced into the opening defined by the bottom part of the retainer ring 27 and is slid down thereon until the projections 29 of the respective legs 28 snap into a bottom circumferential groove 33 of the thimble guide 15. In this manner the retainer ring 27 is axially immobilized on the thimble guide 15 which closely fits into the retainer ring 27. In the absence of the fuel assembly 4 the biased bellows 26 ruges the sleeve 23 of the upper fitting 20 into abutting engagement with the flange 31 of the retainer ring 27 and, at the same time, the bellows 26 ruges the lower fitting 18 downwardly such that its smooth annular radial bottom face 18a surrounding the thimble guide 15 is in an annular contact with the top face 3 of the core plate 2. The fuel assembly 4 is lowered onto the core plate 2 such that the passage 12 of the nozzle platform 7 is in alignment with the passage 22 of the upper fitting 20 of the extender 17. The length of the extender 17 is designed such that before the legs 9 of the fuel assembly nozzle 6 contact the upper face 3 of the core plate 2, the upper annular radial surface area 34 of the cap 21 engages the bottom face 8 of the nozzle platform 7 and, as the fuel assembly 4 continues to move downwardly, the bellows 26 is compressed by virture of the downwardly moving upper fitting 20 causing an increasing, axially oriented resilient force of the bellows 26 to be exerted downwardly on the lower fitting 18 and upwardly on the upper fitting 20. Since the lower fitting 18 is free to slide axially along the ring segments 30 on the lower part of the retainer ring 27, no spring forces are taken up by the retainer ring 27; these forces are applied axially downwardly on the lower fitting 18 further pressing it into engagement with the upper face 3 of the core plate 2. The extender 17 is ready for use as the fuel assembly nozzle 7 assumes its supported position on the core plate 2 as shown in FIG. 1. The upper fitting 20, particularly by virtue of the only narrow annular contact area 25a between the sleeve 23 and the outer face of the retainer ring 27, is capable of executing a slight rocking motion about an axis transverse to the longitudinal extender axis A, readily permitted by the resilient bellows 26. By virtue of such rocking motion any deviation from a strict parallelism between the underside 8 of the nozzle platform 7 and the top face 3 of the core plate 3 and/or between the upper radial annular face 34 of the upper fitting 20 and the bottom annular face 18a of the lower fitting 18 are compensated for and thus face-to-face, sealing engagements between the upper fitting 20 and the platform 7 as well as between the lower fitting 18 and the core plate 2 are ensured. By virtue of the bellows construction and the above-described sealing engagement of the extender 17 with the bottom nozzle 7 and the core plate 2, the inside of the extender 17 is sealed from the surrounding environment, that is, the region between the core plate 2 and the platform 7 of the lower nozzle 6, without the need of accurately fitting, relatively sliding extender parts. Further, by virtue of its sealing contact with the underside 8 of the nozzle platform 7, the cap 21 of the upper fitting 20 seals the passage 12 in the nozzle platform from the surrounding environment and, with its upwardly tapering outer conical face allows the coolant water into platform passages 35 situated adjacent the thimble passage 12. Thus, the thimble extender 17 sufficient shields the thimble 15 from the coolant flow turbulences in the zone between the upper end of the thimble guide 15 and the underface 8 of the nozzle platform 7 and also in the passage 12 and the adjoining instrumentation tube 16. In sealing this area, the extender 17 defines a continuous flow path through the sleeve construction (instrumentation column) 14, the passage 11 and the instrumentation tube 16. By virtue of an orifice (not shown) at the top of the instrumentation tube 16 a reduced flow is obtained through this flow path, resulting, in turn, in smaller forces thus reducing the vibration-causing effect thereof. The cylindrical shield 32 surrounding the bellows 26 protects the bellows from flow turbulences that may cause vibrations thereof and further may be used --by virtue of the cooperation of its upper edge 36 with the underface 37 of teh sleeve 23--to limit the downward axial sliding motion of the upper fitting 20, thus preventing an excessive compression of the bellows 26 particularly during installation of the extender 17 on the thimble guide 15. Such a limiting effect may also be obtained by the cooperation beetween the radial face 21a of the cap 21 and the radial face 31a of the retainer ring 27. In FIG. 3 there is illustrated a thimble guide extender 40 constituting a second preferred embodiment of the invention and shown--similarly to the FIG. 2 embodiment--in its installed state, surrounding thimble guide 15 which is shown in section. The extender 40 comprises a lower fitting generaly designated at 41 having a radially outwardly oriented flange 42 at its upper portion and a cylindrical skirt 43 extending downwardly from the flange 42 and having, at its bottom part, a plurality of circumferentially distributed, radially inwardly projecting resilient locating fingers 44. An upper fitting generally designated at 45 has a cap 46 and a plurality of downwardly extending circumferentially distributed radially resilient fingers 47. The lower and upper fittinsg 41 and 45 are interconnected by a cylindrical, circumferentially closed resilient bellows 48 welded at the top to the lower part of the cap 46 and, at the bottom to the flange 42 of the lower fitting 41. Installation of the thimble guide extender 40 onto the thimble guide 15 is effected in a manner similar to that described in connection with the first embodiment. Thus, the upper conical terminus of the thimble guide 15 is, as the extender 40 is lowered, introduced into the bottom opening defined by the skirt 43 of the lower fitting 412 and the extender 40 is pushed down until the locating fingers 44 snap into the circumferential groove 33 of the thimble guide 15. The inner wall 42a of the flange 42 and the inner wall 43a of the lower part of the skirt 43 engage the respective outer surface portions of the thimble guide 15 with a close, sealing fit and by virtue of the locating fingers 44 engaging into the groove 33, the lower fitting 41 is axially immobilized on the lower part of the thimble guide 15. The sealing engagement with the bottom face 8 of the nozzle platform 7 and the upper annular face of the cap 46 is effected by the resilient force of the bellows 48. In this embodiment, however, the spring force of the bellows 48 is taken up at the bottom by the locating fingers 44 in the groove 33 of the thimble guide 15 and also, if required, by contact between the base of the skirt 43 with the top face 3 of the core plate 2. The lower seal is therefore effected by virtue of the close fit, between the lower fitting 41 and the thimble guide 15, rather than between the top face 3 of the core plate 2 and the lower edge of the skirt 43. The resilient finger 47 of the upper fitting 45 guidingly surrounding the thimble guide 15 permit a rocking motion of the upper fitting 45 about an axis transverse to the longitudinal axis A with the same result and advantages as described in connection with the first embodiment. Also similarly to the first embodiment, the resilient bellows 48 provides a fluidtight closure for the inside of the extender 40 without the need of small tolerances or closely interfitting, operationally sliding extender components. For limiting the extent of compression of the bellows 48, particularly during installation, that is, for limigint the relative motion between the lower fitting 41 and the upper fitting 45, the lower ends 47a of the fingers 47 are in alignment with an upper radial face 49 of the flange 42 of the lower fitting 41 whereby these components are brought into an abutting relationship upon a predetermined motion path which compresses the bellows 48. While in the two described preferred embodiments the sleeve-like sealing unit serves as an extender of and cooperates with a thimble guide mounted on a core plate, it is to be understood that the sleeve-like bellows-equipped sealing unit of the invention may find application in an environment which does not include a thimble guide. It will be understood that the above description of the present invention is susceptible to various modifications, changes and adaptations, and the same are intended to be comprehended within the meaning and range of equivalents of the appended claims. |
044902883 | claims | 1. A method for rapidly removing essentially all of the tritium from a gas mixture, which mixture comprises trace amounts of tritium as contamination, and is contained in an enclosed space, sealed from the atmosphere, said method comprising the steps of (a) removing the gas mixture from said space; (b) passing said gas mixture by means of a blower through a hydrogenating material, comprising an unsaturated carboxylic acid, to remove said contaminating tritium by a hydrogenation reaction with said carboxylic acid; and (c) recirculating the gas obtained from step (b) to said space. 2. The method according to claim 1, which further comprises the step of passing the removed gas mixture from step (a) through an apparatus which converts a given amount of the tritium gas mixture to tritium-containing water, and which separates said water from the remaining tritium gas mixture. 3. The method according to claim 1, wherein the unsaturated carboxylic acid is a polyunsaturated monocarboxylic acid. 4. The method according to claim 3, wherein the polyunsaturated monocarboxylic acid is selected from the group consisting of linoleic acid and linolenic acid. 5. The method according to claim 1, which further comprises adding a catalyst to the hydrogenation reaction. |
description | This application claims priority under 35 U.S.C. §119(e) to U.S. Traditional patent application Ser. No. 11/033,434, filed Jan. 11, 2005 entitled, HELICALLY FLUTED TUBULAR FUEL ROD SUPPORT. 1. Field of the Invention The present invention relates to nuclear reactor fuel assemblies and more particularly to an array for supporting fuel rods wherein the array, or support assembly, consists of a matrix of substantially flat members forming a grid-like frame assembly and a plurality of helically fluted tubular members. 2. Background Information In a typical pressurized water reactor (PWR), the reactor core is comprised of a large number of generally vertically, elongated fuel assemblies. The fuel assemblies include a support grid structured to support a plurality of fuel rods. The fuel assembly includes a top nozzle, a bottom nozzle, a plurality of the support grids and intermediate flow mixing grids, and a plurality of thimble tubes. The support grids are attached to the plurality of elongated thimble tubes which extend vertically between the top and bottom nozzles. The thimble tubes typically receive control rods, plugging devices, or instrumentation therein. A fuel rod includes a nuclear fuel typically clad in a cylindrical metal tube. Generally, water enters the fuel assembly through the bottom nozzle and passes vertically upward through the fuel assembly. As the water passes over the fuel rods, the water is heated until the water exits the top nozzle at a very elevated temperature. The support grids are used to position the fuel rods in the reactor core, resist fuel rod vibration, provide lateral support for the fuel rods and, to some extent, vertically restrain the fuel rods against longitudinal movement. One type of conventional support grid design includes a plurality of interleaved straps that together form an egg-crate configuration having a plurality of roughly square cells which individually accept the fuel rods therein. Depending upon the configuration of the thimble tubes, the thimble tubes can either be received in cells that are sized the same as those that receive fuel rods therein, or can be received in relatively larger thimble cells defined in the interleaved straps. The straps are generally flat, elongated members having a plurality of relatively compliant springs and relatively rigid dimples extending perpendicularly from either side of the flat member. Slots in the straps are utilized to effect an interlocking engagement with adjacent straps, thereby creating a grid of “vertical” and “horizontal” straps which form generally square cells. The location of the springs and dimples are configured such that each cell typically has a spring on each of two adjacent sides. On each of the sides of the cell opposite the springs there are, typically, two dimples. The springs must be disposed opposite the dimples so that the fuel rod is biased against the dimples by the springs. The springs and dimples of each cell engage the respective fuel rod extending through the cell thereby supporting the fuel rod at six points (two springs and four dimples) in each cell. Preferably, each spring and/or dimple includes an arcuate, concave platform having a radius generally the same as a fuel rod. This concave platform helps distribute the radial load on the sides of the fuel rods. The perimeter straps have either springs or dimples extending from one side and peripherally enclose the inner straps of the grid to impart strength and rigidity to the grid. During assembly, the straps must be assembled in a specific configuration to ensure that each cell has the springs and dimples in the proper position. As such, assembly of the prior art frame assembly is a time consuming process. It would be advantageous to have a support assembly that is more easily constructed. The straps may include one or more mixing vanes formed thereon that facilitate mixing of the water within the reactor to promote convective heat exchange between the fuel rods and the water. This motion, along with the elevated temperatures, pressures, and other fluid velocities within the reactor core tend to cause vibrations between the grids and the fuel rods. As with the proper positioning of the straps, care must be used to ensure that the mixing vanes are disposed at the proper locations. Additionally, the action of the water flow impinging on the mixing vanes cause both a pressure drop in the pressure vessel and creates torque in the frame assembly, neither of which are desired. Since the grids support the fuel rods within the fuel cell, such vibrations therebetween can result in fretting of the fuel rods. Such fretting, if sufficiently severe, can result in breach of the fuel rod cladding with resultant nuclear contamination of the water within the reactor. It is desired to provide an improved grid designed to minimize fretting wear between the grids and the fuel rods while maintaining a mixed flow of water through the reactor core. It is also desired to have a support assembly that is easily assembled. There is, therefore, a need to provide a support grid for a nuclear fuel assembly wherein the fuel rods are supported by a tubular member having a helical, fluted fuel rod contact portion. There is a further need for a support assembly that is easily assembled. There is a further need for a nuclear fuel assembly wherein a support grid includes a tubular member having a helical, fluted fuel rod contact portion for supporting fuel rods. These needs, and others, are met by the present invention which provides a support grid for a nuclear fuel assembly, wherein the fuel rod is a generally cylindrical fuel rod with a diameter, and the support grid includes a frame assembly having a plurality of generally uniform cells, each cell having at least one sidewall and a width, and at least one generally cylindrical tubular member. The tubular member has a cell contact portion with a greater diameter and at least one fluted helical fuel rod contact portion with a lesser diameter. As used herein, a “fuel rod contact portion” is typically, but is not limited to, an arcuate line extending at least partly around the cylinder that is a fuel rod. The cell contact portion and the fuel rod contact portion are joined by a transition portion. The greater diameter is generally equivalent to the cell width, and the lesser diameter is generally equivalent to the fuel rod diameter. In this configuration, a fuel rod disposed in the tubular member would engage the inner diameter. The tubular member is disposed in one cell of the plurality of generally square cells so that the cell contact portion engages the at least one cell sidewall. In this manner, the fuel rod is held by the helical fuel rod contact portion and the tubular member is held by the frame assembly. In a preferred embodiment, the tubular member has a wall of uniform thickness so that the helical fuel rod contact portion defines a passage with a helical shape on both the side adjacent to the fuel rod and the side adjacent to the cell wall. These helical shaped passages act to mix the water so that mixing vanes are not required. There are at least two advantages to using the helical shaped passages; first, the water flow does not impinge on the shaped passage, so there is a minimal pressure drop created by the mixing structure. Second, by reversing the direction of the helical passage in selected cells, the amount of torque exerted on the frame assembly may be controlled. The helical fuel rod contact portion may be formed in various configurations. For example, there may be a single (or multiple) helical fuel rod contact portion having an angular displacement of 360 degrees, that is, extending 360 degrees around the tubular member. However, given the relatively short height of a typical cell, the pitch (radial distance/height) of the helical fuel rod contact portion may be too great thereby restricting the flow of water through the helical portion of the passage. Alternatively, there may be at least two helical fuel rod contact portions each extending 180 degrees around the tubular member. However, in a preferred embodiment, there are four helical fuel rod contact portions each extending 90 degrees around the tubular member. While these examples have used a number (N) of helical fuel rod contact portions and an angular displacement (A) that equals 360 (N*A=360), this is not required. That is, virtually any number of helical fuel rod contact portion(s) may be used with any angular displacement. It is further noted that, while a symmetrical helical contact portion is preferred, a helical contact portion may be an asymmetrical helix; that is the pitch may be variable along the tubular member. The tubular members, preferably, have a smooth transition between the cell contact portion and the helical fuel rod contact portion. Where there are four helical fuel rod contact portions, the shape of the tubular member is similar to the perimeter of a flower with four petals. Alternatively, the tubular member may include extended platform sections structured to engage either the wall of the frame assembly and/or the fuel rod. Where there is a platform, the transition section will typically be a sharp curve. In another embodiment, the greater portion of the length of the transition portion is generally flat and the ends are sharply angled. The frame assembly includes a plurality of cells typically structured to contain a nuclear fuel rod. As noted above, some cells are adapted to enclose a thimble rod or other device. However, the non-fuel rod cells are not relevant to this invention and, while noted, will not be discussed hereinafter. In the preferred embodiment, the frame assembly is made from a plurality of substantially flat, elongated strap members disposed in two interlocked sets, a “vertical” set and a “horizontal” set. The vertical set of strap members is disposed generally perpendicular to the horizontal strap members. Also, the strap members in each set are generally evenly spaced. In this configuration, the cells are generally square. In an alternate embodiment, the frame assembly is made from tubular members that have been welded together, preferably at 90 degree intervals. As used herein, directional terms, such as, but not limited to, “upper” and “lower” relate to the components as shown in the Figures and are not limiting upon the claims. As shown in FIG. 1, there is a fuel assembly 20 for a nuclear reactor. The fuel assembly 20 is disposed in a water vessel (not shown) having an inlet at the bottom and an outlet at the top. The fuel assembly 20 comprises a lower end structure or bottom nozzle 22 for supporting the fuel assembly 20 on the lower core plate (not shown) in the core region of a reactor (not shown); a number of longitudinally extending control rod guide tubes, or thimbles 24, projecting upwardly from the bottom nozzle 22; a plurality of transverse support grids 26 axially spaced along the guide thimbles 24; an organized array of elongated fuel rods 28 transversely spaced and supported by the grids 26; an instrumentation tube 30 located in the center of the assembly; and an upper end structure or top nozzle 32 attached to the upper ends of the guide thimbles 24, in a conventional manner, to form an integral assembly capable of being conventionally handled without damaging the assembly components. The bottom nozzle 22 and the top nozzle 32 have end plates (not shown) with flow openings (not shown) for the upward longitudinal flow of a fluid coolant, such as water, to pass up and along the various fuel rods 28 to receive the thermal energy therefrom. To promote mixing of the coolant among the fuel rods 28, a mixing vane grid structure, generally designated by the numeral 34, is disposed between a pair of support grids 26 and mounted on the guide thimbles 24. The top nozzle 32 includes a transversely extending adapter plate (not shown) having upstanding sidewalls secured to the peripheral edges thereof in defining an enclosure or housing. An annular flange (not shown) is secured to the top of the sidewalls. Suitably clamped to this flange are leaf springs 36 (only one of which being shown in FIG. 1) which cooperate with the upper core plate (not shown) in a conventional manner to prevent hydraulic lifting of the fuel assembly caused by upward coolant flow while allowing for changes in fuel assembly length due to core induced thermal expansion and the like. Disposed within the opening defined by the sidewalls of the top nozzle 32 is a conventional rod cluster control assembly 38 having radially extending flukes, being connected to the upper end of the control rods, for vertically moving the control rods in the control rod guide thimbles 24 in a well known manner. To form the fuel assembly 20, support grids 26 and a mixing vane grid structure 34 are attached to the longitudinally extending guide thimbles 24 at predetermined axially spaced locations. The bottom nozzle 22 is suitably attached to the lower ends of the guide thimbles 24 and then the top nozzle 32 is attached to the upper ends of guide thimbles 24. Fuel rods 28 are then inserted through the grids 26 and grid structure 34. The fuel rods 28 are generally elongated cylinders having a diameter. For a more detailed description of the fuel assembly 20, reference should be made to U.S. Pat. No. 4,061,536. The fuel assembly 20 depicted in the drawings is of the type having a square array of fuel rods 28 with the control rod guide thimbles 24 being strategically arranged within the fuel rod array. Further, the bottom nozzle 22, the top nozzle 32, and likewise the support grids 26 are generally square in cross section. In that the specific fuel assembly 20 represented in the drawings is for illustrational purposes only, it is to be understood that neither the shape of the nozzles or the grids, or the number and configuration of the fuel rods 28 and guide thimbles 24 are to be limiting, and the invention is equally applicable to different shapes, configurations, and arrangements than the ones specifically shown. For example, as shown in FIGS. 2 and 4, the support grid 26 includes a frame assembly 40 and at least one generally cylindrical tubular member 50. The frame assembly 40 includes a plurality of cells 42 defined by cell walls 43. Each cell 42 has a width as indicated by the letter “w.” In one embodiment, the cells 42 and cell walls 43 are formed from a plurality of substantially flat, elongated strap members 44 disposed in two interlocked sets, a vertical set 46 and a horizontal set 48. The strap members 44 in the vertical and horizontal sets 48 of strap members 44 are generally perpendicular to each other. Additionally, the strap members 44 in each set are generally evenly spaced. In this configuration, the strap members 44 form generally square cells 42A. Thus, each cell 42A has two diagonal axes “d1” and “d2,” which are perpendicular to each other and extend through the corners of the cell 42A, as well as two normal axes “n1” and “n2,” which are perpendicular to each other and extend through the center of the cell 42A and which intersect perpendicularly with the cell walls 43. The points on the cell wall 43 that the two normal axes pass through are the closest point, “cp,” between the cell wall 43 and the center of the cell 42. As shown in FIG. 3, the frame assembly 40 also has a height, indicated by the letter “h,” wherein the height is substantially less than the width or length of the frame assembly 40. Further, the frame assembly 40 has a top side 47 and a bottom side 49. It is notable that the strap members 44 of the present invention do not include protuberances, such as springs and dimples, as did strap members of the prior art. The lack of additional support structures make the construction of the frame assembly 40 very easy. The tubular member 50 of the support grid 26 is shown in FIGS. 4 and 5. The tubular member 50 includes at least one helical fluted portion or fuel rod contact portion 52, a cell contact portion 54, and a transition portion 56 disposed therebetween. As shown in FIGS. 4-6, the tubular member 50 has four fuel rod contact portions 52, which is the preferred embodiment. Other configurations are discussed below. The cell contact portion 54 has a greater diameter being generally equivalent to said cell width and is structured to snugly engage the cell 42. The fuel rod contact portion 52 has a lesser diameter, being generally equivalent to said fuel rod 28 diameter. Thus, the tubular member 50 may be disposed in a cell 42 and a fuel rod 28 may be disposed in the tubular member 50. In a preferred embodiment, the tubular member 50 is made from a material having a uniform thickness. Thus, the helical fuel rod contact portion 52 defines an outer passage 60 between the outer side of the tubular member 50 and the cell wall 43. Additionally, the cell contact portion 54, which is spaced from the fuel rod 28, defines an inner passage 62. Water which flows through either the outer or inner passages 60, 62 is influenced by the shape of the helical fuel rod contact portion 52 resulting in the water being mixed. The tubular member 50 may be constructed with any number of helical fuel rod contact portions 52 which may have any degree of pitch. For example, as shown in FIG. 7, a tubular member 50 has a single helical fuel rod contact portion 52 that extends 360 degrees about the tubular member 50. As shown in FIG. 8, a tubular member 50 has a two helical fuel rod contact portions 52 that each extend 180 degrees about the tubular member 50. As shown in FIG. 9, a tubular member 50 has a two helical fuel rod contact portions 52 that each extend 360 degrees about the tubular member 50. As noted above, FIG. 5 shows a tubular member 50 having a four helical fuel rod contact portions 52 that each extend 90 degrees about the tubular member 50. Preferably, the helical fuel rod contact portions 52 are spaced evenly about the tubular member 50, but this is not required. These examples have used a number (N) of helical fuel rod contact portions 52 and an angular displacement (A) that equals 360 degrees or a multiple of 360 degrees. This configuration is especially adapted for use in a square cell 42A. That is, the cell contact portion 54 will only contact the cell wall 43 at the closest point on the cell wall 43. At other points, e.g., the corner of the cell 42A, the tubular member 50 greater diameter, that is the cell contact portion 54, will not contact a cell wall 43. Thus, as shown best in FIG. 6, where there are four evenly spaced, helical fuel rod contact portions 52 that each extend 90 degrees about the tubular member 50, there are four corresponding cell contact portions 54, each disposed between a helical fuel rod contact portions 52. To ensure the greatest amount of surface area contact between the tubular member 50 and the cell wall 43, the tubular member 50 is disposed with each helical fuel rod contact portion 52 generally aligned with a diagonal axis at the top side 47 of the cell and aligned with a different diagonal axis at the bottom side 49 of the cell. In this orientation, the cell contact portion 54 is aligned with a cell wall 43 closest point at the top side 47 and at the bottom side 49. A similar configuration may be made with cells 42 of any shape. That is, the number (N) of helical fuel rod contact portions 52 is preferably equal to the number of sides (S) to the cell 42, and the angular displacement (A) is preferably 360 degrees/S. Thus, the tubular member may be positioned with each helical fuel rod contact portion 52 generally aligned with an axis passing through the corner of the cell 42 at the top side 47 of the cell and aligned with a different axis passing through the corner of the cell 42 at the bottom side 49 of the cell. Thus, the cell contact portion 54 is aligned with the cell wall 43 closest point at the top side 47 and at the bottom side 49. In another embodiment, the frame assembly 40 includes a plurality of cylindrical cells 42B defined by a plurality of connected tubular frame members 70. As shown in FIG. 10, the frame assembly 40 may have a plurality of densely packed tubular frame members 70, however, as shown in FIG. 11, a pattern of aligned tubular frame members 70 is preferred. That is, the tubular frame members 70 are coupled to each other at 90 degree intervals about the perimeter of each tubular frame member 70. The tubular member 50 is disposed within the cylindrical cells 42B. As shown in FIG. 12, the combination of the tubular member 50 and the cylindrical cell 42B again creates an inner passage 62 between the fuel rod 28 and the tubular member 50 and an outer passage 60 between the tubular member 50 and the tubular frame member 70. The cylindrical cell 42B of the tubular frame member 70 has the additional advantage that the entire cell contact portion 54 abuts the cell wall 43. That is, the diameter of the cylindrical cell 42B is the same as the cell width, which is also the same as the closest point, and, as such, the cell contact portion 54 will engage the cell wall 43 along the entire height of the cell wall 43. This is unlike a square cell 42A wherein the cell contact portion 54 does not contact the cell wall 43 at the corners. In another embodiment, shown in FIG. 13, the functions of the tubular member 50 and the tubular frame member 70 have been combined in a helical frame member 80. That is, the frame assembly 40 includes a plurality of helical frame members 81 disposed in a matrix pattern. The helical frame member 80, like the tubular member 50, includes at least one helical fuel rod contact portion 52, however, instead of a cell contact portion 54, the outer side of the helical frame member 80 is a contact portion 55 structured to be directly coupled to the contact portion 55 of an adjacent helical frame member 80. As with the tubular frame member 70 embodiment of the frame assembly 40, the helical frame members 80 are coupled to each other at 90 degree intervals about the perimeter of each helical frame member 80. Additionally, in this embodiment the frame assembly 40 preferably includes a plurality of outer straps 82 structured to extend about the perimeter of the plurality of helical frame members 81. The outer straps 82 are coupled to the contact portion 55 of the helical frame members 80 disposed at the outer edge of the plurality of helical frame members 81. A fuel rod 28 is disposed through at least one helical frame member 80. As shown best in FIG. 12, as viewed as a cross-section, the tubular member 50 components, i.e., the helical fuel rod contact portion 52, the cell contact portion 54, and the transition portion 56, preferably, are shaped as smooth curves. This configuration gives the tubular member 50 a compressible, spring-like quality. However, as shown in FIG. 14, the cell contact portion 54 may include an extended planar length or platform 90. The platform 90 is structured to provide a greater surface area which engages the cell wall 43. The greater length of the platform 90 will necessitate the transition portion 56 having a sharp curve. Similarly, as shown in FIG. 15, the helical fuel rod contact portion 52 may include a concave platform 92 adapted to extend radially about the fuel rod 28. As before, greater length of the concave platform 92 will necessitate the transition portion 56 having a sharp curve. A tubular member 50 may also include both a platform 90 at the cell contact portion 54 and a concave platform 92 at the helical fuel rod contact portion 52. Finally, the tubular member 50 may also be constructed with a generally flat transition portion 56 with angled ends 94. As shown in FIG. 16, in this embodiment the transition portion 56 is generally planar in a cross-sectional top view. It is understood that, due to the helical nature of the fuel rod contact portion 52, the transition portion 56 is not flat in the direction of the height of the frame assembly 40. While specific embodiments of the invention have been described in detail, it will be appreciated by those skilled in the art that various modifications and alternatives to those details could be developed in light of the overall teachings of the disclosure. Accordingly, the particular arrangements disclosed are meant to be illustrative only and not limiting as to the scope of invention which is to be given the full breadth of the claims appended and any and all equivalents thereof. |
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summary | ||
abstract | A differential phase contrast X-ray imaging system includes an X-ray illumination system, a beam splitter arranged in an optical path of the X-ray illumination system, and a detection system arranged in an optical path to detect X-rays after passing through the beam splitter. |
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063114763 | summary | TECHNICAL FIELD The present invention relates generally to power generating systems for orbit transfer vehicles. More particularly, the present invention relates to a power generating system that is incorporated with a direct gain propulsion system to convert solar thermal energy into electric energy that may be stored in a battery. BACKGROUND OF THE INVENTION Background Art Space propulsion systems using solar thermal energy have been proposed as a means to achieve greater payload fractions. These engines can be used, for example, to provide thrust to boost payloads from low earth orbits to higher orbits or to alter the orbit of a payload. In such engines, solar radiation is captured and focused by mirrors into a "black body" where the solar radiation is used to heat a propellant, such as hydrogen. The propellant is then passed through a nozzle to create thrust. To further reduce the weight of the orbit transfer vehicle and increase the payload fraction, U.S. Pat. No. 5,459,996 entitled "Hybrid Solar Rocket Utilizing Thermal Storage for Propulsion and Electrical Power", the disclosure of which is hereby incorporated by reference as if fully set forth herein, has suggested a solar thermal rocket engine that utilizes energy conversion diodes to convert stored thermal energy into electrical energy. This configuration, however, does not reduce the weight of the orbit transfer vehicle to the maximum extent possible due to the relative inefficiency with which stored thermal energy is converted into electrical energy. The inefficiency of this process necessitates the use of a large and relatively heavy thermal storage device. SUMMARY OF THE INVENTION It is one object of the present invention to provide a solar thermal engine which does not rely on stored thermal energy for powering a craft. It is a more specific object of the present invention to provide a solar thermal engine which employs a plurality of static power converters to convert ambient thermal energy to electrical energy which is then stored in an electrical storage device. It is yet another object of the present invention to provide a method for propelling and powering a craft which does not rely on stored thermal energy for powering the craft. In one preferred form, the present invention provides a solar thermal engine for propelling and powering a craft. The solar thermal engine includes a housing, a propellant annulus, a plurality of static power converters and an electrical energy storage device. The housing has an optical cavity for receiving a beam of concentrated sunlight and converting the beam into ambient thermal energy. The propellant annulus is coupled to the housing and is selectively operable in a heating mode wherein the propellant annulus transmits at least a first portion of the ambient thermal energy to heat a flow of propellant. The plurality of static power converters are coupled to the housing and receive the first portion of the ambient thermal energy when the propellant annulus is not operated in the heating mode. The plurality of static power converters employ the first portion of the ambient thermal energy to generate electrical energy. The electrical energy storage device is coupled to the plurality of static power converters and receives and stores the electrical energy generated by the plurality of static power converters. A method for propelling and powering a craft is also provided. |
052805085 | summary | This invention relates to the insertion and replacement of fuel rods to fuel bundles. More particularly, and in the case of fuel bundles having part length rods, provision is made for the insertion and replacement of recessed part length rods to enable inspection and or replacement of the part length fuel rods without disturbing or requiring disassembly of the rest of the fuel bundle. BACKGROUND OF THE INVENTION In Dix et al. U.S. Pat. No. 5,112,570 entitled Two Phase Pressure Drop Reduction BWR Assembly Design, the concept of partial length rods is disclosed. This concept may best be understood by first understanding the construction of a conventional fuel bundle with conventional fuel rods and thereafter setting forth in summary format the part length rod invention of the above referenced patent, which patent is incorporated by reference in this disclosure. A conventional boiling water fuel bundle includes a lower tie plate for supporting a matrix of vertically upstanding fuel rods at their lower end and an upper tie plate for holding the same matrix of vertical upstanding sealed fuel rods at their upper end. The lower tie plate permits the inflow of liquid moderator (water) around the fuel rods, the upper tie plate permits the outflow of liquid moderator (water) and generated vapor moderator (steam) at the upper end of the fuel bundle. A channel surrounds the lower tie plate, surrounds the upper tie plate, and defines a confined flow path for the moderator between the upper and lower tie plates about the matrix of upstanding fuel rods. Fluid flow is thus confined between the upper and lower tie plates and isolated from surrounding liquid moderator in the so-called core by pass area immediate adjoined to the fuel channel on the outside of the fuel channel. The fuel rods within the fuel bundle are flexible and unless otherwise restrained would come into contact with one another under the forces of flow induced vibration and so-called "creep"--a differential growth in the fuel rods resulting from their heated, pressurized and radioactive environment. This being the case, a system of fuel rod spacers is distributed from the top to the bottom of the fuel bundles. These spacers form a matrix of individual cells discretely surrounding each fuel rod at spaced apart elevations within the fuel bundle holding the flexible fuel rods in their designed side-by-side relationship. During operation of the fuel bundle in a reactor core of a boiling water nuclear reactor, the fuel bundle can be dynamically described as having two regions of operation. These regions include a lower single phase region containing liquid moderator (water) and an upper two phase region containing liquid moderator (water) intermixed with increasing fractions of vapor moderator (steam). In a boiling water nuclear reactor, the moderator serves two purposes. First, the moderator moderates fast neutrons generated by the reaction into slow or thermal neutrons necessary to continue the reaction. Secondly, the dense water moderator turns to expanded saturated steam. Energy is extracted from the saturated steam by passing the steam through an engine, such as a steam turbine. Having set forth the conventional construction and operation of a boiling water nuclear reactor fuel bundle, the concept of part length rods can now be discussed. In the above reference Dix et al. U.S. Pat. No. 5,112,570, so-called part length rods were disclosed. In short, an invention is set forth in which "a plurality of fuel rods extending from (the) lower tie plate toward the upper tie plate , (the) part length rods terminating within the two phase region of the bundle before reaching the upper tie plate;". The invention makes the point that "at least two of the part length rods (are) separate from one another so as to define in at least two locations in (the) bundle spaced apart and separate vents commencing at the top of said partial length rods and extending to (the) upper tie plate." Specifically, "each of (the) spaced apart vents (is) immediately adjoined by adjacent full length fuel rods." Usefulness of the invention is set forth. Specifically, an improved fuel to moderator ratio is created in the upper two phase region of the fuel bundle. More importantly, pressure drop reduction in the upper two phase region of the fuel bundle is set forth. This enables greater stability of the fuel bundle and a reactor core including fuel bundles against thermal hydraulic and nuclear, thermal hydraulic instabilities. Additionally, the fuel to moderator ratio is improved, especially in the cold operating state. The reader will understand at this point that the disclosed part length rods relate to the required spacers in two ways. First, since the part length rods do not extend to the upper tie plate, it is the spacers that hold the part length rods vertically upright at their respective upper ends. Secondly, since the part length rods terminate before the upper tie plate, some of the spacers in the upper two phase region of the fuel bundle overlying the ends of the part length rods. As will hereinafter appear, it is these spacers that constitute an obstacle in the desired removal of the part length rods. It is common to inspect fuel bundles, and especially the individual fuel rods of fuel bundles during the operational life times of the fuel bundles. Unfortunately, the very presence of the part length rods renders the inspection of the part length rods inherently difficult. A brief understanding of the constraints of such inspections can be helpful. Inspections of fuel bundle parts are typically made during reactor outages. During such reactor outages, the power output of the reactor is lost. This loss of power outage carries with it a corresponding loss in revenue. Any delay prolonging the reactor outage can be costly--running into lost revenues of well over several hundred thousand dollars per hour! Accordingly, provision must be made for rapid inspection of all parts of a fuel bundle, including the new part length rods. Conventional inspection of fuel bundles is typically accomplished in a submerged environment within a so-called "holding pool." The fuel bundle removed from the reactor is placed upright within the holding pool. Thereafter, the channel surrounding the fuel rods and the upper tie plate holding the fuel rods are removed. In the case of the fuel bundle having nothing but full length rods, the individual fuel rods may thereafter be accessed at the exposed top of each fuel rod, individually removed, inspected and replaced. Access to the fuel rods occurs at the top of the fuel rods. Such access is a routine matter. Unfortunately, part length rods present special problems. First, the part length rods have a length from the lower tie plate that is less than the full length rods. For example, the typical full length fuel rod is in the order of 160 inches in length; the typical part length fuel rod is in the order of 120 inches in length. In order to reach the part length rod, one has to penetrate a matrix of full length fuel rods. No view of the engagement of the part length rod is possible. All engagement is essentially "blind." Secondly, since the part length fuel rods do not extend to the upper tie plate, the part length fuel rods are held in their upright position by the spacers. Accordingly, any tool for the removal and replacement of the part length fuel rods can have a diameter dimension no greater than the diameter dimension of the fuel rods being inspected. It is required that the tool for the removal of the part length rods pass through the spacers overlying the part length rods. Finally, it is required in some applications that the part length rods be fastened against vertical movement to the lower tie plate. This is done by screwing the part length rods into receiving threads on the lower tie plate. Accordingly, the disassembly of such part length rods is correspondingly rendered more complicated. The screw threads stick. The part length fuel rod must be forcibly rotated. Accordingly, there is a need for both a tip on the part length rod and corresponding tools to render the rapid replacement and removal of such part length rods during reactor outages reliable and fast. SUMMARY OF THE PRIOR ART In the testing of the concept set forth in Dix et al. U.S. Pat. No. 5,112,570, entitled Two Phase Pressure Drop Reduction BWR Assembly Design, lead test assemblies were utilized having part length rods in certain locations of the fuel bundle. In the case of these lead test assemblies, the part length rods were not secured to the lower tie plate by a threaded connection; the part length rods rested with their weight being utilized to provided the connection to the lower tie plate. These lead test assemblies required the removal of the part length rods for inspection. In that case, each part length rod was equipped with a cylindrical tip. The cylindrical tip had a female screw driver indentation on the top and was provided with external male threads. Removal of the part length rods from the interior of the fuel bundle occurred by tool having an exterior tube with corresponding female threads and an internal, freely rotating male screw driver. In operation, the idea was to prevent the rotation of the part length rod through engagement of the male screw driver head with threaded engagement of the female threads of the extraction tool with the male threads of the part length rod tip. This prior art provision for the removal of the part length rods from the lead test assemblies was less than optimum. Considerable difficulty was experience with the registration of the male screw driver fitting of the extraction tool with the female screw driver indentation upwardly exposed from the tip. Further, threading of the male threads on the tip with the female threads of the tool was also less than optimum. The tool was pendulously suspended from a bridge a considerable distance to the particular part length rod to be remove and inspected. As a result, the female threads of the suspended tube were vertical. The part length rod, however, had been subjected to the forces of rod creep--a non linear differential growth--which often left the protruding threads of the part length rod less than vertical. Threading blind with the particular threads perfectly vertically aligned was difficult. The added complexity of the exposed threads of the part length rods being less than vertical made the part length rod removal task exceeding difficult in many cases. The reader will realize that the particular apparatus for the removal of the part length rods is prior art. The discovery of the difficulty in the intended removal of the part length rods is not prior art. This difficulty was only discovered when actual removal was attempted and is not known in the art. SUMMARY OF THE INVENTION A part length fuel rod tip and group of tip grasping tools is set forth for the removal, inspection and replacement of part length fuel rods from a fuel bundle having a part length rods interspersed with a majority of conventional full length rods. The part length fuel rod tip includes a longitudinal keyway allowing torque to be exerted on the fuel rod and a horizontal slot enabling grasping of the fuel rod for vertical withdrawal and replacement movement of the fuel rod. The tools include a torque socket for applying high torque forces to the part length rod for unscrewing the rod when the rod becomes stuck at its threaded connection to the lower tie plate, a tip grasping tool for permitting normal unscrewing torque and grasping for vertical withdrawal and replacement, and finally a tool having both high torque and grasping characteristics. In each case, the diameter of the tools is restricted to enable access to the part length fuel rods through the spacers overlying the upwardly exposed ends of the part length fuel rods. OTHER OBJECTS, FEATURES AND ADVANTAGES It is a first object of this invention to disclose a tip for a part length fuel rod which enables rotation under high torque forces as well as vertical withdrawal, manipulation for inspection, and finally enabling convenient replacement of the part length fuel rod. The tip constitutes a generally solid cylindrical addition to the fuel bundle constructed from the same material as the cladding of the fuel bundle--usually the alloy known as zircaloy. The upper surfaces of the tip are rounded, imparting male gathering surfaces to the upwardly exposed end of the fuel rod. The tip includes at least one longitudinally extending keyway. This keyway enables the tip--and the fuel rod to which it is fixed--to be rotated under forces of high torque. Finally, at a portion of the tip relatively close to the fuel rod, a female cylindrical section is removed from the tip S to define a stop surface. This surface enables the tip to be firmly grasped to assure both rotation under high torque and vertical withdrawal and insertion of the fuel rod. Referring to the "female cylindrical section" removed from the tip S, it will be understood that those having acquaintance with the mechanical design arts would preferred to use other descriptive terms. This portion of the tip could be referred to as a "horizontal slot", "horizontal keyway", "fly cut notch", and other similar terms. We use the term "female cylindrical section" so that this specification can be geometrically descriptive for the purposes of describing this invention with precision. An advantage of the disclosed tip is that it does not interfere with the conventional portions of the tip design including provisions for charging the tip with gas under pressure to suppress gas discharge from the fission process occurring within the fuel rod. Stopping at this point, the reader must remember the circumstances of the required inspection of the part length rods. During the reactor outage, time is of the essence. Further, and due to the fact that the part length rods are radio active and therefore must be handled with relative rapidity under water in a holding pool, manipulation of the following disclosed tools occurs at the end of a long pole within an upright fuel bundle having its channel and upper tie plate removed. Finally, the pole must penetrate into the matrix of full length rods--and water rods--upwardly exposed in the pool. At the bottom of the pole--out of view of the personnel performing the inspection--all manipulation of the part length rod must occur. An additional object of this invention is to disclose an alternate tool for rotating the part length fuel rod under high torque. According to this aspect of the invention, a female socket is disclosed. The socket include an inside bore having an inside diameter equal to the outside diameter of the part length rod tip. This inside bore includes a protruding key. This key is dimensioned with respect to the bore and spaced with respect to the bore to fit interiorly of the tip keyway. By remote rotation of the tool, remote rotation under high torque can be exerted on the part length rod for unscrewing motion from the lower tie plate. Another object of this invention is to disclose a general purpose tool for the normal manipulation of a part length rod. According to this aspect of the invention, a female tool member is disclosed having a slit peripheral cylinder defining at least three discrete cylinder segments. The slits in the cylinder defining the cylinder segments are longitudinal and at preferable 120.degree. intervals about the cylinder. This cylinder is given a dimension enabling the cylinder segments to be separated overlying the part length rod tip so as to fit in a snug and slightly expanded relationship over the tip outer surface. A remotely actuated reciprocating sleeve moves to and from a surrounding and embracing relationship over the slit cylinder. When the sleeve is withdrawn from the slit cylinder, insertion and withdrawal of the sleeve from the partial length rod tip can occur. When the sleeve overlies the slit cylinder, locking of the sleeve to the part length fuel rod tip at the outer surface only occurs. In this locked state, normal torque forces as well as vertical forces of withdrawal and insertion can be exerted on the part length fuel rod through the tip. An additional object of this invention is to disclose a tool which locks to the keyway and cylindrical section of the partial length fuel rod tip. According to this aspect of the invention, a tool is disclosed having a spring biased tang with an attached key and male cylinder segment for fitting to the female cylinder segment of the tip. This tang fits to a tool shaft in a tang slot with the key and male cylinder section spring biased into a tip locking position at that end of the tang remote from the point of attachment to the shaft. A reciprocating sleeve operates over the tang sprung key and male cylinder section. In operation, when the sleeve is withdrawn, the tool may be placed over the tip, rotated until spring biased engagement of the key at the longitudinal keyway occurs, and thereafter advanced onto the tip until the male cylinder segment of the tool engages in and locks to the female cylinder segment in the tip. When the sleeve is advanced overlying the tang, key and male cylinder segment, locking of the tool at the tip occurs enabling rotation under relatively high torque as well as axial replacement and withdrawal of the part length fuel rod. Finally, a ramp is provided at the upward end of the tang attached male cylinder segment to enable climbing of the male cylinder segment of the tool free of the female cylinder segment of the tip to enable tool removal from the tip. An advantage of the spring biased key is that it permits relative rotation and tactile location of the keyway at the end of the part length rod from the required remote location of personnel inspecting the part length fuel rod. A further advantage of all disclosed tool constructions is that they can freely pass to and from the part length fuel rod tips without disturbing the surround full length fuel rods or (more importantly) the overlying spacers. As a result, inspection can be conveniently carried out in the short time confines of a reactor outage. |
description | This application is based upon and claims the benefit of priority from the prior Japanese Patent Application No. 2013-121717 filed on Jun. 10, 2013 in Japan, the entire contents of which are incorporated herein by reference. 1. Field of the Invention The present invention relates to a multi charged particle beam writing method and a multi charged particle beam writing apparatus. More specifically, for example, the present invention relates to a blanking method in writing with multiple beams. 2. Description of Related Art The lithography technique that advances miniaturization of semiconductor devices is extremely important as being a unique process whereby patterns are formed in the semiconductor manufacturing. In recent years, with high integration of LSI, the line width (critical dimension) required for semiconductor device circuits is decreasing year by year. The electron beam (EB) writing technique, which intrinsically has excellent resolution, is used for writing or “drawing” a pattern on a wafer and the like with electron beams. As an example employing the electron beam writing technique, a writing apparatus using multiple beams (multi-beams) can be cited. Compared with the case of writing a pattern by using a single electron beam, since it is possible to emit multiple beams at a time in multiple writing, the throughput can be greatly increased. In the writing apparatus employing a multi-beam system, for example, multiple beams are formed by letting an electron beam emitted from an electron gun assembly pass through a mask with a plurality of holes, blanking control is performed for each of the beams, and each unblocked beam is reduced by an optical system and deflected by a deflector so as to irradiate a desired position on a target object or “sample” (refer to, e.g., Japanese Patent Application Laid-open (JP-A) No. 2006-261342). In the multi-beam writing, the dose of an individual beam is individually controlled by an irradiation time. For highly accurately controlling the dose of each beam, it is necessary to carry out blanking control at high speed to perform a beam ON/OFF control. Conventionally, in a writing apparatus of a multi-beam system, a blanking control circuit for each beam is placed on a blanking plate where each blanking electrode of multiple beams is arranged. Controlling is asynchronously performed for each beam. For example, a trigger signal for causing a beam to be ON is sent to control circuits of all the beams. In responsive to the trigger signal, the control circuit of each beam applies a beam-on voltage to an electrode and, simultaneously, starts counting the irradiation time period by a counter. Then, when the irradiation time has been completed, a beam-off voltage is applied. In performing such a control, a 10-bit control signal has been used, for example. However, since the space for placing a circuit on a blanking plate and the amount of current to be used are restricted, there is no other alternative but to have an uncomplicated circuit for the amount of information of control signals. Therefore, it has been difficult to build in a blanking circuit that can perform an operation of high speed and high precision. Further, installing a blanking control circuit for each beam on a blanking plate restricts to narrow the pitch of multiple beams. By contrast, when placing a control circuit for each beam outside the writing apparatus body and connecting each of them by wiring in order to secure a space for installing the circuit, since the wiring becomes long, there is a problem that a crosstalk problem becomes more prominent. In accordance with one aspect of the present invention, a multi charged particle beam writing method includes converting, for each shot of beams of multiple charged particle beams, a respective first gray scale value of each beam of the multiple charged particle beams, which is obtained by dividing an individual irradiation time period of the each beam by a quantization unit, into respective data of binary numbers of a predetermined digit number set in advance; dividing a maximum irradiation time period per shot of beams of the multiple charged particle beams into a plurality, being the predetermined digit number, of first irradiation time periods, each of which is calculated by multiplying a corresponding second gray scale value of a plurality of second gray scale values by the quantization unit, where the plurality of second gray scale values are gray scale values defined in decimal numbers converted from each digit value of data of binary numbers of the predetermined digit number; dividing a plurality of second irradiation time periods, which are a part of the plurality of first irradiation time periods into a plurality of third irradiation time periods; dividing irradiation of each beam of the multiple charged particle beams by using the plurality of third irradiation time periods and remaining undivided plurality of first irradiation time periods, into the first irradiation steps which are irradiation steps of the plurality of third irradiation time periods and second irradiation steps which are irradiation steps of the remaining undivided plurality of first irradiation time periods; and irradiating, for each group of a plurality of groups in the each shot of beams, a target object, in order, with the multiple charged particle beams such that the plurality of groups are respectively composed of combination of at least two irradiation steps of first irradiation steps and second irradiation steps and the plurality of groups continue in order. In accordance with another aspect of the present invention, a multi charged particle beam writing apparatus includes a stage configured to mount a target object thereon and to be continuously movable; an emission unit configured to emit a charged particle beam; an aperture member, in which a plurality of openings are formed, configured to form multiple beams by letting a region including a whole of the plurality of openings be irradiated with the charged particle beam and letting portions of the charged particle beam respectively pass through a corresponding opening of the plurality of openings; a plurality of blankers configured to respectively perform blanking deflection of a corresponding beam in the multiple beams having passed through the plurality of openings of the aperture member; a blanking aperture member configured to block each beam having been deflected to be in a beam-off state by the plurality of blankers; and a deflection control unit configured to control a corresponding blanker of the plurality of blankers such that a maximum irradiation time period per shot of beams of the multiple beams is divided into a plurality, being a predetermined digit number set in advance, of first irradiation time periods, each of which is calculated by multiplying a corresponding second gray scale value of a plurality of second gray scale values by a quantization unit, where the plurality of second gray scale values are gray scale values defined in decimal numbers converted from each digit value of data of binary numbers of the predetermined digit number, a plurality of second irradiation time periods, which are a part of the plurality of first irradiation time periods, are divided into a plurality of third irradiation time periods, irradiation of each beam of the multiple charged particle beams by using the plurality of third irradiation time periods and remaining undivided plurality of first irradiation time periods is divided into first irradiation steps of the plurality of third irradiation time periods and second irradiation steps of the remaining undivided plurality of first irradiation time periods, and a target object is irradiated, in order, with the multiple charged particle beams such that the plurality of groups are respectively composed of combination of at least two irradiation steps of the first irradiation steps and the second irradiation steps and the plurality of groups continue in order, for each group of a plurality of groups in the each shot of beams. An irradiation method has been examined in which irradiation of each shot of beams is divided into a plurality of irradiation steps obtained by converting an irradiation time of each shot of beams into binary numbers, defining a binary number of each digit in a decimal number to be equivalent to an irradiation time of each digit, and combining the irradiation time of each digit to be digit number irradiation steps, and then, two-digit grouping is sequentially performed by combining a smaller irradiation time and a longer irradiation time to execute irradiation in order of the group. However, a problem has arisen in the method described above that, since there is a great difference between totals of irradiation time of groups, when performing irradiation of a group which is next to a group whose total irradiation time is extremely short, data transmission may not follow the irradiation operation of the group whose total irradiation time is extremely short, and thereby the data transmission time may be a latency time for a beam irradiation operation. Therefore, a further improvement is needed. Then, in the following Embodiments, there will be described a writing apparatus and method that can reduce or avoid the latency time for a beam irradiation operation, due to data transmission time, while maintaining restriction of a circuit installation space. In the following embodiments, there will be described a configuration in which an electron beam is used as an example of a charged particle beam. The charged particle beam is not limited to the electron beam, and other charged particle beam such as an ion beam may also be used. FIG. 1 is a schematic diagram showing a configuration of a writing apparatus according to the first embodiment. In FIG. 1, a writing (or “drawing”) apparatus 100 includes a writing unit 150 and a control unit 160. The writing apparatus 100 is an example of a multi charged particle beam writing apparatus. The writing unit 150 includes an electron optical column 102 and a writing chamber 103. In the electron optical column 102, there are arranged an electron gun assembly 201, an illumination lens 202, an aperture member 203, a blanking plate 204, a reducing lens 205, a deflector 212, a limiting aperture member 206, an objective lens 207, and a deflector 208. In the writing chamber 103, there is arranged an XY stage 105. On the XY stage 105, there is placed a target object or “sample” 101 such as a mask serving as a writing target substrate when performing writing. The target object 101 is, for example, an exposure mask used for manufacturing semiconductor devices, or a semiconductor substrate (silicon wafer) on which semiconductor elements are formed. The target object 101 may be, for example, a mask blank on which resist is applied and a pattern has not yet been formed. On the XY stage 105, further, there is arranged a mirror 210 for measuring the position of the XY stage. The control unit 160 includes a control computer 110, a memory 112, a deflection control circuit 130, a logic circuit 132, a stage position measurement unit 139, and storage devices 140, 142, and 144, such as magnetic disk drives. The control computer 110, the memory 112, the deflection control circuit 130, the stage position measurement unit 139, and the storage devices 140, 142, and 144 are mutually connected through a bus (not shown). Writing data is input into the storage device 140 (storage unit) from the outside to be stored therein. In the control computer 110, there are arranged an area density calculation unit 60, an irradiation time calculation unit 62, a gray scale value calculation unit 64, a bit conversion unit 66, a bit processing unit 70, a writing control unit 72, a bit processing table generation unit 73, an exposure table generation unit 74, and a transmission processing unit 68. Each function, such as the area density calculation unit 60, the irradiation time calculation unit 62, the gray scale value calculation unit 64, the bit conversion unit 66, the bit processing unit 70, the writing control unit 72, the bit processing table generation unit 73, the exposure table generation unit 74, or the transmission processing unit 68 may be configured by hardware such as an electronic circuit, or by software such as a program implementing these functions. Alternatively, they may be configured by a combination of hardware and software. Data which is input and output to/from the area density calculation unit 60, the irradiation time calculation unit 62, the gray scale value calculation unit 64, the bit conversion unit 66, the bit processing unit 70, the writing control unit 72, the bit processing table generation unit 73, the exposure table generation unit 74, or the transmission processing unit 68, and data being calculated are stored in the memory 112 each time. FIG. 1 shows a structure necessary for explaining the first embodiment. Other structure elements generally necessary for the writing apparatus 100 may also be included. FIGS. 2A and 2B are conceptual diagrams each showing an example of the configuration of an aperture member according to the first embodiment. In FIG. 2A, holes (openings) 22 are formed at a predetermined arrangement pitch, in the shape of a matrix, in the aperture member 203, wherein m×n (m≧2, n≧2) holes 22 are arranged in m columns in the vertical direction (the y direction) and n rows in the horizontal direction (the x direction). In FIG. 2A, holes 22 of 512 (rows)×8 (columns) are formed, for example. Each hole 22 is a quadrangle of the same dimensions and shape. Alternatively, each hole may be a circle of the same circumference. In this case, there is shown an example of each row having eight holes 22 from A to H in the x direction. Multi-beams 20 are formed by letting portions of an electron beam 200 respectively pass through a corresponding hole of a plurality of holes 22. Here, there is shown the case where the holes 22 are arranged in two or more columns and rows in both the x and the y directions, but it is not limited thereto. For example, it is also acceptable to arrange a plurality of holes 22 in only one row or in only one column, that is, in one row where a plurality of holes are arranged as columns, or in one column where a plurality of holes are arranged as rows. Moreover, the method of arranging the holes 22 is not limited to the case of FIG. 2A where holes are aligned in a grid. It is also preferable to arrange the holes 22 as shown in FIG. 2B where the position of each hole in the second row is shifted from the position of each hole in the first row by a dimension “a” in the horizontal direction (x direction), for example. Similarly, it is also preferable to arrange the holes 22 such that the position of each hole in the third row is shifted from the position of each hole in the second row by a dimension “b” in the horizontal direction (x direction). FIG. 3 is a conceptual diagram showing the configuration of a blanking plate according to the first embodiment. FIG. 4 is a top view conceptual diagram showing the configuration of a blanking plate according to the first embodiment. In the blanking plate 204, a passage hole is formed to be corresponding to the arrangement position of each hole 22 of the aperture member 203, and a pair of electrodes 24 and 26 (blanker: blanking deflector) is arranged for each passage hole. An amplifier 46 for applying voltage is respectively arranged at one (for example, the electrode 24) of the two electrodes 24 and 26 for each beam. A logic circuit 41 is independently arranged at the amplifier 46 for each beam respectively. The other one (for example, the electrode 26) of the two electrodes 24 and 26 for each beam is grounded. An electron beam 20 passing through a corresponding passage hole is respectively deflected by the voltage applied to the two electrodes 24 and 26 being a pair. Blanking control is performed by this deflection. Thus, a plurality of blankers respectively perform blanking deflection of a corresponding beam in the multiple beams having passed through a plurality of holes 22 (openings) of the aperture member 203. FIG. 5 is a schematic diagram showing the internal configuration of an individual blanking control circuit and a common blanking control circuit according to the first embodiment. In FIG. 5, a shift register 40, a register 42, a selector 48, and an AND computing unit 44 (logical product computing unit) are arranged in each logic circuit 41 for individual blanking control arranged at the blanking plate 204 in the body of the writing apparatus 100. The AND computing unit 44 may be omitted. According to the first embodiment, for example, a 2-bit control signal is used for individual blanking control for each beam, which has conventionally been controlled by, for example, a 10-bit control signal. That is, for example, a 2-bit control signal is input/output to/from the shift register 40, the register 42, the selector 48 and the AND computing unit 44. Since the amount of information of a control signal is small, an installation area of the control circuit can be small. In other words, even when a logic circuit is arranged on the blanking plate 204 whose installation space is small, more beams can be arranged at a smaller beam pitch. This enables the amount of current passing the blanking plate to be increased, and therefore, a writing throughput can be improved. Moreover, an amplifier is arranged at the deflector 212 for common blanking, and a register 50 and a counter 52 are arranged at the logic circuit 132. These do not perform several different controlling at the same time, and therefore, it is sufficient to use one circuit to perform ON/OFF control. Accordingly, even when arranging a circuit for a high speed response, no problem occurs with respect to the restriction on the installation space and the current to be used in the circuit. Therefore, this amplifier is operated at very high speed compared with an amplifier realizable on a blanking aperture. This amplifier is controlled by a 10-bit control signal, for example. That is, for example, a 10-bit control signal is input/output to/from the register 50 and the counter 52. According to the first embodiment, blanking control of each beam is performed by using both the beam ON/OFF control by each logic circuit 41 for individual blanking control described above and the beam ON/OFF control by the logic circuit 132 for common blanking control that collectively control all the multiple beams. FIG. 6 is a flowchart showing main steps of a writing method according to the first embodiment. In FIG. 6, a series of steps: a pattern area density calculation step (S102), a shot time period (irradiation time) T calculation step (S104), a gray scale value N calculation step (S106), a conversion to binary number step (S108), an irradiation time arrangement data processing step (S109), an irradiation time arrangement data output step (S110), a target group data transmission step (S112), a writing step (S114) based on irradiation time of a target group, a determination step (S120), a group change step (S122), and a determination step (S124) are executed. The writing step (S114) based on irradiation time of a target group executes, as its internal steps, a series of steps: an individual beam ON/OFF switching step (S116) and a common beam ON/OFF switching step (S118). In the pattern area density calculation step (S102), the area density calculation unit 60 reads writing data from the storage device 140, and calculates the area density of a pattern arranged in the writing region of the target object 101 or in each mesh region of a plurality of mesh regions made by virtually dividing a chip region to be written into meshes. For example, the writing region of the target object 101 or a chip region to be written is divided into strip-shaped stripe regions each having a predetermined width. Then, each stripe region is virtually divided into a plurality of mesh regions described above. It is preferable that the size of a mesh region is, for example, a beam size, or smaller than a beam size. For example, the size of a mesh region is preferably about 10 nm. The size density calculation unit 60 reads corresponding writing data from the storage device 140 for each stripe region, and allocates a plurality of figure patterns defined in the writing data to a mesh region, for example. Then, the area density of a figure pattern arranged in each mesh region is to be calculated. In the shot time period (irradiation time) T calculation step (S104), the irradiation time calculation unit 62 calculates an irradiation time T (which hereinafter will also be called a shot time period or an exposure time) of the electron beam per shot, for each predetermined sized mesh region. When performing multi-pass writing, an irradiation time T of the electron beam per shot in each hierarchy of multi-pass writing is to be calculated. It is preferable to obtain an irradiation time T, being a reference, to be in proportion to the area density of a calculated pattern. Moreover, it is preferable that the irradiation time T to be finally calculated is a time equivalent to a dose after correction, that is a dose having been corrected with respect to a dimension change amount for a phenomenon causing dimension variations, such as a proximity effect, a fogging effect, or a loading effect not shown. The size of a plurality of mesh regions for defining the irradiation time T and the size of a plurality of mesh regions where a pattern area density is defined may be the same size or different sizes. When they are different sizes, each irradiation time T is calculated after interpolating an area density by linear interpolation, etc. The irradiation time T for each mesh region is defined in an irradiation time map, and the irradiation time map is stored in the storage device 142, for example. In the gray scale value N calculation step (S106), the gray scale value calculation unit 64 calculates a gray scale value N, being an integer, which is used when defining the irradiation time T for each mesh region, defined in the irradiation time map, by using a predetermined quantization unit Δ. The irradiation time T is defined by the following equation (1).T=ΔN (1) Therefore, the gray scale value N is defined as an integer value obtained by dividing the irradiation time T by a quantization unit Δ. The quantization unit Δ can be variously set, and, for example, can be defined by 1 ns (nanosecond), etc. It is preferable that a value of 1 to 10 ns, for example, is used as the quantization unit Δ. Δ indicates a quantization unit for controlling, such as a clock period, in the case of performing control by a counter. In the conversion to binary number step (S108), the bit conversion unit 66 converts, for each shot of beams of multiple beams, a gray scale value N (first gray scale value) of each beam of the multiple beams, which is obtained by dividing an irradiation time (individual irradiation time period) of each of multiple beams by the quantization unit Δ, into a binary value (respective data of binary numbers) of n-digit set in advance. For example, when N=50, since it is 50=21+24+25, if converting into a 10-digit binary value, it becomes “0000110010”. For example, if N=500, it is “0111110100”. For example, if N=700, it is “1010111100”. For example, if N=1023, it is “1111111111”. For each shot, the irradiation time of each beam is equivalent to an irradiation time defined for a mesh region to be irradiated by each beam concerned. Thereby, the irradiation time T is defined by the following equation (2). T = Δ ∑ k = 0 n - 1 a k 2 k ( 2 ) ak indicates a value (1 or 0) of each digit in the case defining the gray scale value N by a binary number. Although it is sufficient for n, being the number of digits, to be two or more, preferably it should be four or more digits, and more preferably, it should be eight or more digits. According to the first embodiment, for each shot of beams, irradiation of each beam of each shot of beams concerned is divided into irradiation steps of “n” times, “n” being the number of digits of a binary number sequence (data of binary numbers) set in advance. The irradiation steps of n times is equivalent to a combination of irradiation of irradiation time periods (plurality of first irradiation time periods). A maximum irradiation time period per shot of beams of the multiple beams is divided into a plurality, being the digit number “n”, of the irradiation time periods (plurality of first irradiation time periods). Each of the irradiation time periods (plurality of first irradiation time periods) is calculated by multiplying a corresponding gray scale value (second gray scale value) of a plurality of gray scale values (plurality of second gray scale values) by Δ, where the plurality of gray scale values (plurality of second gray scale values) are gray scale values defined in decimal numbers converted from each digit value of a binary value (data of binary numbers) of n-digit. In other words, one shot of a beam is divided into a plurality of irradiation steps of irradiation time periods of Δa020, Δa121, . . . , Δak2k, . . . , Δan-12n-1. In the case of n=10, n being the number of digits, one shot is divided into ten irradiation steps. FIG. 7 shows a bit processing table representing a relation between each digit number and an irradiation time of each digit in the case of the digit number n=10 according to the first embodiment. In FIG. 7, the irradiation time of the first digit (k=0) (the first bit) is Δ, the irradiation time of the second digit (k=1) (the second bit) is 2Δ, the irradiation time of the third digit (k=2) (the third bit) is 4Δ, the irradiation time of the fourth digit (k=3) (the fourth bit) is 8Δ, . . . , and the irradiation time of the tenth digit (k=9) (the tenth bit) is 512Δ. For example, in the case of the digit number n being 10 (n=10), if N=70, the irradiation time of the tenth digit (the tenth bit) is Δ×512. The irradiation time of the ninth digit (the ninth bit) is Δ×0=0. The irradiation time of the eighth digit (the eighth bit) is Δ×128. The irradiation time of the seventh digit (the seventh bit) is Δ×0=0. The irradiation time of the sixth digit (the sixth bit) is Δ×32. The irradiation time of the fifth digit (the fifth bit) is Δ×16. The irradiation time of the fourth digit (the fourth bit) is Δ×8. The irradiation time of the third digit (the third bit) is Δ×4. The irradiation time of the second digit (the second bit) is Δ×0=0. The irradiation time of the first digit (the first bit) is Δ×0=0. The total time of these is 700Δ. For example, in the case of performing irradiation in order from the largest digit to the smallest digit, if Δ=1 ns, the first irradiation step is irradiation of 512 ns (beam ON). The second irradiation step is irradiation of 0 ns (beam OFF). The third irradiation step is irradiation of 128 ns (beam ON). The fourth irradiation step is irradiation of 0 ns (beam OFF). The fifth irradiation step is irradiation of 32 ns (beam ON). The sixth irradiation step is irradiation of 16 ns (beam ON). The seventh irradiation step is irradiation of 8 ns (beam ON). The eighth irradiation step is irradiation of 4 ns (beam ON). The ninth irradiation step is irradiation of 0 ns (beam OFF). The tenth irradiation step is irradiation of 0 ns (beam OFF). There has been described the case of transmitting data for “n” irradiation steps in order of the amount of data from the largest, for example. The time for data transmission can be included in the irradiation time of an irradiation step by performing in parallel the transmission of data indicating ON/OFF of the (k−1)th bit (the (k−1)th digit)) of each beam with the irradiation step of the k-th bit (the k-th digit) of each beam. However, if k becomes small, since the irradiation time of an irradiation step becomes short, it is difficult to include the transmission of data indicating ON/OFF of the (k−1)th bit (the (k−1)th digit)) in the irradiation time of the irradiation step. Then, a digit whose irradiation time is long and a digit whose irradiation time is short are grouped. Thereby, the data transmission time of the next group can be included in the total of grouped irradiation time in the irradiation step. FIG. 8 shows a grouped exposure table according to a comparative example of the first embodiment. FIG. 8 shows the case of n=10 similarly to FIG. 7. In the example of FIG. 8, in order that the difference between the totals of grouped irradiation time may become smaller to be close to uniform, configuration is performed as follows. As the exposure step 1, the group 1 is composed of the first digit (k=0) (the first bit) and the tenth digit (k=9) (the tenth bit) of the bit processing table of FIG. 7. As the exposure step 2, the group 2 is composed of the second digit (k=1) (the second bit) and the ninth digit (k=8) (the ninth bit). As the exposure step 3, the group 3 is composed of the third digit (k=2) (the third bit) and the eighth digit (k=7) (the eighth bit). As the exposure step 4, the group 4 is composed of the fourth digit (k=3) (the fourth bit) and the seventh digit (k=6) (the seventh bit). As the exposure step 5, the group 5 is composed of the fifth digit (k=4) (the fifth bit) and the sixth digit (k=5) (the sixth bit). Thus, as described above, by dividing the irradiation time into five groups, the difference between the totals of the grouped irradiation time can be small compared with the case of no grouping. However, as shown in FIG. 8, the total of the irradiation time of the group 1 shown as the exposure step 1 is 513Δ, whereas, the total of the irradiation time of the group 5 shown as the exposure step 5 is 48Δ. Thus, there is a difference of ten times or more between the totals of the exposure time (irradiation time) of the exposure steps, which means that there still exists a large difference. For example, in the case of the total of the irradiation time of the group 5 shown as the exposure step 5 being 48Δ, it is necessary to increase the operation clock of the shift register 40 so that data transmission of the group to be exposed next may be completed within while the irradiation of the group 5 shown as the exposure step 5 is being performed. Then, according to the first embodiment, irradiation time of some digits of the bit processing table shown in FIG. 7 is divided further. FIGS. 9A and 9B show internal configuration of the bit processing table generation unit and the exposure table generation unit according to the first embodiment. As shown in FIG. 9A, in the bit processing table generation unit 73, there are arranged an initial setting unit 80, a reference irradiation time T′ calculation unit 82, a determination unit 84, an irradiation time increase number “a” change unit 86, and a dividing unit 88. As shown in FIG. 9B, in the exposure table generation unit 74, there are arranged an assignment processing unit 90 and an adjustment unit 92. Each function, such as the initial setting unit 80, the reference irradiation time T′ calculation unit 82, the determination unit 84, the irradiation time increase number “a” change unit 86, the dividing unit 88, the assignment processing unit 90 and the adjustment unit 92 may be configured by hardware such as an electronic circuit or by software such as a program causing a computer to implement these functions. Alternatively, it may be configured by a combination of hardware and software. Data which is input and output to/from the initial setting unit 80, the reference irradiation time T′ calculation unit 82, the determination unit 84, the irradiation time increase number “a” change unit 86, the dividing unit 88, the assignment processing unit 90 and the adjustment unit 92 and data being calculated are stored in the memory 112 each time. FIG. 10 is a flowchart showing a generation method of a bit processing table and an exposure table according to the first embodiment. In FIG. 10, the generation method of a bit processing table and an exposure table executes a series of steps: an initial setting step (S20), a reference irradiation time T′ calculation step (S22), a determination step (S24), an irradiation time increase number “a” change step (S26), a dividing step (S30), a grouping processing step (S32), and a time adjustment step (S34). The time adjustment step (S34) may be omitted. The bit processing table generation unit 73 generates a bit processing table showing a relation between a place value “k” of bit data and an irradiation time corresponding to the place value “k” by the following calculation procedure. A maximum irradiation time per shot of beams of the multiple beams is divided into “n” number of plurality of irradiation time periods (first irradiation time periods). Each of the irradiation time periods (plurality of first irradiation time periods) is calculated by multiplying a corresponding gray scale value (second gray scale value) of a plurality of gray scale values (plurality of second gray scale values) by Δ, where the plurality of gray scale values (plurality of second gray scale values) are gray scale values defined in decimal numbers converted from each digit value of a binary value (data of binary numbers) of n-digit. Then, a part (second irradiation time periods), being a number “b”, of the “n” plurality of irradiation time periods is further divided into a plurality of irradiation time periods (third irradiation time periods). Then, divided (a+b) plurality of irradiation time periods (third irradiation time periods), “a” being explained below, and the remaining undivided (n−b) plurality of irradiation time periods (first irradiation time periods) are used for generating a bit processing table. In the initial setting step (S20), the initial setting unit 80 sets an initial value for each of a combination number “m” and an increased number “a” of irradiation time periods (an irradiation time increased number “a”), wherein the increased number “a” of irradiation time periods indicates that the number of irradiation time periods is increased by “a”. Since 1 bit is necessary for irradiation time arrangement data of the irradiation step of each digit, for example, if when configuring data transmission by 2-bit data, since it becomes combination (grouping) of irradiation steps of two digits, the combination number “m” is two (m=2). For example, if when configuring data transmission by 3-bit data, since it becomes combination (grouping) of irradiation steps of three digits, the combination number “m” is three (m=3). For example, if when configuring data transmission by 4-bit data, since it becomes combination (grouping) of irradiation steps of four digits, the combination number “m” is four (m=4). Here, for example, it is supposed that “m” is two (m=2). For example, in the list shown in FIG. 7, the irradiation time is divided into ten irradiation time periods, where ten is the number of digits whose place values “k” are 0 to 9. For example, if dividing two of the ten irradiation time periods into four, since two irradiation time periods become four irradiation time periods, the ten-digit number irradiation time periods become twelve irradiation time periods totally, and thus, the irradiation time increased number “a” is two (a=2). For example, if dividing two of the ten irradiation time periods into six, since two irradiation time periods become six irradiation time periods, the ten-digit irradiation time periods become fourteen irradiation time periods totally, and thus, the irradiation time increased number “a” is four (a=4). In this case, it is supposed that “a” is two (a=2). In the reference irradiation time T′ calculation step (S22), the reference irradiation time T′ calculation unit 82 calculates a reference irradiation time T′ by solving the following equation (3), using the combination number “m”, the digit number “n”, the irradiation time increased number “a” and the quantization unit Δ. T ′ = ( 2 n - 1 ) ( n + a ) / m Δ ( 3 ) For example, in the case of n=10, m=2, and a=2, the reference irradiation time T′=170.5Δ (=1023Δ/{(10+2)/2}). Then, it is determined whether the calculated reference irradiation time T′ is appropriate or not. In the determination step (S24), the determination unit 84 determines whether the calculated reference irradiation time T′ satisfies the following equation (4) or not, by using the irradiation time increased number “a”, the irradiation time Ti (first irradiation time period) of the i-th digit of the binary number in “n” irradiation time periods (first irradiation time periods), where “n” being the number of digits, and “b” being the number of irradiation time periods to be divided as a part of the “n” irradiation time periods (first irradiation time periods). T ′ > ∑ i Ti > T ′ Ti a + b ( 4 ) For example, in the case of n=10, m=2, a=2, and the reference irradiation time T′=170.5Δ in the bit table of FIG. 7, with respect to the irradiation time period Ti that exceeds 170.5Δ, there are two 256Δ and 512Δ. According to the first embodiment, irradiation time periods exceeding the reference irradiation time T′ are dividing targets. Therefore, “b”, which is the number of irradiation time periods (second irradiation time periods) to be divided, as a part of the “n” irradiation time periods (first irradiation time periods) is calculated to be two (b=2). Accordingly, the right-hand side of the equation (4) is calculated to be (256+512)/(2+2)=192, which does not satisfy the equation (4). When the calculated reference irradiation time T′ does not satisfy the equation (4), it proceeds to the irradiation time increase number “a” change step (S26). In the irradiation time increase number “a” change step (S26), the irradiation time increase number “a” change unit 86 changes an increased number “a” of irradiation time periods. Here, it is changed to a=4, for example. Then, it returns to the reference irradiation time T′ calculation step (S22). Then, in the determination step (S24), each of the steps from the reference irradiation time T′ calculation step (S22) to the irradiation time increase number “a” change step (S26) is repeated until the calculated reference irradiation time T′ satisfies the equation (4). In the reference irradiation time T′ calculation step (S22) after the increased number “a” of irradiation time periods has been changed to a=4, for example, the equation (3) is calculated similarly. For example, in the case of n=10, m=2, and a=4, the reference irradiation time is T′=146.1Δ. Next, it is determined whether the calculated reference irradiation time T′=146.1Δ is appropriate or not. In the determination step (S24), with respect to the irradiation time period that exceeds 146.1Δ, there are two 256Δ and 512Δ. Therefore, “b” is two (b=2). However, since a=4, the right-hand side of the equation (4) is calculated to be (256+512)/(4+2)=128. Therefore, the calculated reference irradiation time T′=146.1Δ satisfies the equation (4). Accordingly, the reference irradiation time T′=146.1Δ is appropriate, and, in this regard, the number “b” of the irradiation time periods to be divided is two (b=2), and the irradiation time increase number “a” is four (a=4). As described above, the reference irradiation time T′, and the number “b” of irradiation time periods to be divided and the increased number “a” of irradiation time periods, concerning this reference irradiation time T′, are calculated. In the determination step (S24), when the calculated reference irradiation time T′ satisfies the equation (4), it proceeds to the dividing step (S30). In the dividing step (S30), with respect to the “n” irradiation time periods (first irradiation time periods), “n” being the number of digits, the dividing unit 88 divides “b” irradiation time periods Ti (second irradiation time periods), each of which is greater than the reference irradiation time T′, into a plurality of irradiation time periods (third irradiation time periods) so that the number of a plurality of irradiation time periods (first irradiation time periods) may be increased by “a”. Specifically, for example, in the above-described case of n=10, m=2, a=4, b=2, and T′=146.1Δ, there are two irradiation time periods Ti, namely 256Δ and 512Δ. Therefore, the two irradiation time periods, 256Δ and 512Δ, are divided into six (a+b) irradiation time periods. In that case, it is preferable to use the irradiation time of the i-th digit of the binary digit which is the closest to the reference irradiation time T′, as an object of the irradiation time to be divided. Here, 128Δ is the closest to T′=146.1Δ. Therefore, the two irradiation time periods 256Δ and 512Δ are divided into six irradiation time periods each being 128Δ. FIG. 11 shows a bit processing table representing a relation between each digit number and an irradiation time of each digit, after the dividing, in the case of the digit number n=10 according to the first embodiment. In FIG. 11, the irradiation time from the first digit (k=0) (the first bit) to the eighth digit (k=7) (the eighth bit) are the same as those in FIG. 7. In FIG. 11, the ninth digit (k=8) (the ninth bit) is divided into k=8a and k=8b, the irradiation time of each of which is made to be 128Δ. The tenth digit (k=9) (the tenth bit) is divided into k=9a, 9b, 9c, and 9d, the irradiation time of each of which is made to be 128Δ. Thus, combination of fourteen (n+a) irradiation time periods is totally obtained. As described above, the bit processing table generation unit 73 generates a bit processing table which is for generating (n+a)-digit binary number data to define an irradiation time per shot. The generated bit processing table is stored in the storage device 144. The ON/OFF data of irradiation time after the dividing is configured so that the ON/OFF data of irradiation time before dividing may be succeeded. That is, for example, if the ON/OFF data of the irradiation time of the tenth digit (the tenth bit) is ON, the ON/OFF data of the irradiation time of the divided k=9a, 9b, 9c and 9d is also to be ON. If the ON/OFF data of the irradiation time of the ninth (the ninth bit) is ON, the ON/OFF data of the irradiation time of the divided k=8a and k=8b is also to be ON. Thereby, even if the dividing is performed, the total of the irradiation time per shot can be the same. The bit processing table should be generated before starting writing processing. By the procedure described above, “n” irradiation time periods, “n” being the digit number, are regenerated to be (n+a) irradiation time periods. In other words, one shot is redivided into (n+a) irradiation steps from “n” irradiation steps. Next, the exposure table generation unit 74 generates a grouped exposure table by assigning each irradiation time of the generated bit processing table to one of a plurality of groups (irradiation time group) which is composed a combination of at least two irradiation time periods. In the grouping processing step (S32), for each shot of beams, the assignment processing unit 90 performs assignment by the following calculation procedure. As mentioned above, a maximum irradiation time period per shot of beams of the multiple beams is divided into “n” irradiation time periods (a plurality of first irradiation time periods). Each of the “n” irradiation time periods is calculated by multiplying a corresponding gray scale value (second gray scale value) of a plurality of gray scale values (plurality of second gray scale values) by Δ, where the plurality of gray scale values (plurality of second gray scale values) are gray scale values defined in decimal numbers converted from each digit value of a binary value (data of binary numbers) of n-digit. Further, “b” irradiation time periods (plurality of second irradiation time periods), which are a part of the “n” irradiation time periods, are divided into “(b+a)” irradiation time periods (plurality of third irradiation time periods). The assignment processing unit 90 assigns the plurality of irradiation time periods (third irradiation time periods) and the remaining undivided irradiation time periods (first irradiation time periods) to one of a plurality of irradiation time groups which is composed a combination of at least two irradiation time periods. Specifically, the assignment is performed as follows. The assignment processing unit 90 assigns the divided irradiation time periods (third irradiation time periods) and the remaining undivided irradiation time periods (first irradiation time periods) to one of a plurality of groups so that the total irradiation time of each group may further be close to the reference irradiation time T′. Here, combination of a smaller (shorter) irradiation time and a larger (longer) irradiation time is assigned in order. In other word, irradiation of each beam of the multiple charged particle beams is divided into first irradiation steps of the divided irradiation time periods and second irradiation steps of the remaining undivided irradiation time periods, and a plurality of groups are respectively composed of combination of at least two irradiation steps of the first irradiation steps and the second irradiation steps for each beam of each shot of beams. FIG. 12 shows a grouped exposure table according to the first embodiment. Similarly to FIG. 11, FIG. 12 shows the case where the irradiation time (exposure time) is divided into fourteen irradiation time periods. In the example of FIG. 12, in order that the difference between totals of grouped irradiation time may become smaller to be close to uniform, configuration is performed as follows. As the exposure step 1, the group 1 is configured by the first digit (k=0) (the first bit) and the fourteenth digit (the fourteenth bit) which is a division (k=9d) of the tenth digit (the tenth bit) in the bit processing table of FIG. 11. As the exposure step 2, the group 2 is configured by the second digit (k=1) (the second bit) and the thirteenth digit (the thirteenth bit) which is a division (k=9c) of the tenth digit (the tenth bit). As the exposure step 3, the group 3 is configured by the third digit (k=2) (the third bit) and the twelfth digit (the twelfth bit) which is a division (k=9b) of the tenth digit (the tenth bit). As the exposure step 4, the group 4 is configured by the fourth digit (k=3) (the fourth bit) and the eleventh digit (the eleventh bit) which is a division (k=9a) of the tenth digit (the tenth bit). As the exposure step 5, the group 5 is configured by the fifth digit (k=4) (the fifth bit) and the tenth digit (the tenth bit) which is a division (k=8b) of the ninth digit (the ninth bit). As the exposure step 6, the group 6 is configured by the sixth digit (k=5) (the sixth bit) and the ninth digit (the ninth bit) which is a division (k=8a) of the ninth digit (the ninth bit). As the exposure step 7, the group 7 is configured by the seventh digit (k=6) (the seventh bit) and the eighth digit (k=7) (the eighth bit). Thus, by dividing into seven groups, the difference between the totals of the irradiation time of the groups can be small compared with the case of the five groups of FIG. 8. In the comparison example of FIG. 8, there is a difference of ten times or more between the totals of the exposure time (irradiation time) of exposure steps. On the other hand, according to the first embodiment, as shown in FIG. 12, the total of the irradiation time of the group 1 shown as the exposure step 1 is 129Δ, whereas the total of the irradiation time of the group 7 shown as the exposure step 7 is 192Δ. Thus, the difference of the total of exposure time (irradiation time) can be reduced to 1.49 times between exposure steps. Accordingly, when performing data transmission processing of the group 1 shown as the exposure step 1 whose irradiation time is the shortest, it is sufficient just to increase the operation clock of the shift register 40 to be one and a half times, and thus, it is unnecessary to increase it to be ten times as described in the comparative example of FIG. 8. The grouped exposure table generated as described above is stored in the storage device 144. The exposure table should be generated before starting writing processing. Although, in the example described above, the grouped exposure table is generated in the writing apparatus 100, it is not limited thereto. If the digit number “n” which is used when converting the irradiation time per shot into binary number data has been previously set, the grouped exposure table itself can also be set beforehand. Therefore, it is also preferable to generate a grouped exposure table outside the apparatus in advance, and to input it to the writing apparatus 100 to be stored in the storage device 144. In other words, it is also preferable to prepare the bit processing table generation unit shown in FIGS. 9A and 9B, as an external device. If further reducing the difference between totals of the exposure time (the irradiation time) of exposure steps, what is necessary is just to execute the time adjustment step (S34). In addition, the time adjustment step (S34) may be omitted. In the time adjustment step (S34), in order to make the difference between the total irradiation time periods of a plurality of groups be closer to each other, the adjustment unit 92 divides partial irradiation time of each irradiation time configuring some groups of a plurality of groups into a plurality of irradiation time periods (fourth irradiation time periods), and assigns one irradiation time period of the plurality of irradiation time periods (fourth irradiation time periods) to other group. In order to make the total irradiation time of each group be closer to the reference irradiation time T′, the adjustment unit 92 divides partial irradiation time of each irradiation time configuring some groups into a plurality of irradiation time periods (fourth irradiation time periods), and assigns one irradiation time period of the plurality of irradiation time periods (fourth irradiation time periods) to other group. FIG. 13 shows an example of a grouped exposure table after the adjustment according to the first embodiment. In FIG. 13, a part of irradiation time of a group whose total irradiation time is the largest is divided. In the example of FIG. 12, the largest of the total irradiation time is 192Δ of the group 7 shown as the exposure step 7. The reference irradiation time T′ in this example is T′=146.1Δ. Therefore, the difference between them is about 46Δ. Accordingly, it is desired to make the total of the irradiation time of the group 7 shown as the exposure step 7 be 146Δ. However, the group 7 shown as the exposure step 7 is composed of the irradiation time 64Δ of the seventh digit (k=6) (the seventh bit) and the irradiation time 128Δ of the eighth digit (k=7) (the eighth bit), and therefore, is not a dividing target irradiation time to be divided. If dividing is performed for an irradiation time other than a dividing target irradiation time to be divided, b=2 described above cannot be obtained. Then, according to the first embodiment, adjustment is performed using a dividing target irradiation time. Specifically, for example, in the case of n=10, m=2, a=4, b=2, and T′=146.1Δ as described above, the number of exposure steps (the number of groups) is seven as described above. Therefore, seven irradiation time periods in order from the smallest are separately assigned to respective exposure steps (groups) in order to increase possibility of making each of the seven irradiation time periods and any one of irradiation time periods divided from at least one dividing target irradiation time be respectively included in a same step of the exposure steps (groups). Next, the adjustment unit 92 calculates a combination to be closer to the reference irradiation time T′=146.1Δ in the case of combining irradiation time periods which are not dividing targets. Since the remaining irradiation time not being a dividing target is the eighth digit (k=7) (the eighth bit) irradiation time 128Δ, it is possible to obtained the total of irradiation time 144Δ by combining it with 16Δ of the exposure step 5 (the group 5) in FIG. 13. Next, with respect to the group 7 shown as the exposure step 7 whose total of irradiation time is 192Δ was the maximum in the case of FIG. 12, since the irradiation time 64Δ of the seventh digit (k=6) (the seventh bit) has already been assigned, the remaining 82Δ is obtained by dividing the irradiation time 128Δ of the division (k=9d) of the tenth digit (the tenth bit) into 82Δ and 46Δ, and then assigning the 82Δ. Thereby, the total of the irradiation time of the group 7 shown as the exposure step 7 can be 146Δ. Next, the group 1 is configured by the first digit (k=0) (the first bit) and the division (k=9a) of the tenth digit (the tenth bit). As the exposure step 2, the group 2 is configured by the second digit (k=1) (the second bit) and the division (k=9b) of the tenth digit (the tenth bit). As the exposure step 3, the group 3 is configured by the third digit (k=2) (the third bit) and the division (k=9c) of the tenth (tenth bit). Since k=0 is Δ, k=1 is 2Δ, and k=2 is 4Δ, they are all small. Then, the remaining 46Δ of the divided k=9d is divided into 16 Δ, 16Δ, and 14Δ, and assigned to the irradiation time of k=9a, 9b and 9c. Thereby, with respect to the group 1 shown as the exposure step 1, the total of the irradiation time can be 145Δ. With respect to the group 2 shown as the exposure step 2, the total of the irradiation time can be 146Δ. With respect to the group 3 shown as the exposure step 3, the total of the irradiation time can be 146Δ. After the processing described above, remaining groups are the group 4 shown as the exposure step 4 and the group 6 shown as the exposure step 6. As the exposure step 4, the group 4 is configured by the fourth digit (k=3) (the fourth bit) and the division (k=8a) of the ninth digit (the ninth bit). As the exposure step 6, the group 6 is configured by the sixth digit (k=5) (the fifth bit) and the division (k=8b) of the ninth digit (the ninth bit). Here, with respect to the group 4 shown as the exposure step 4, since the irradiation time 8Δ of the fourth digit (k=3) (the fourth bit) and 128Δ of k=8a have been assigned, it is necessary to add 10Δ. On the other hand, with respect to the group 6 shown as the exposure step 6, since the irradiation time 32Δ of the sixth digit (k=5) (the sixth bit) and 128Δ of k=8b have been assigned, 14Δ is superfluous. Then, 128Δ of k=8b is divided into 118Δ and 10Δ. 118Δ is assigned to t8b of the group 6 shown as the exposure step 6, and 10Δ is assigned to t8a of the group 4 shown as the exposure step 4. Thereby, the total of the irradiation time of the group 4 shown as the exposure step 4 can be 146Δ. The total of the irradiation time the group 6 shown as the exposure step 6 can be 150Δ. As described above, by performing the time adjustment step (S34), the respective total irradiation time periods of a plurality of groups can be closer to each other. In addition, the ON/OFF data of the irradiation time after the adjustment is configured so that the ON/OFF data of the irradiation time before dividing may be succeeded. Therefore, it is impossible to make an element of an exposure table by using each element of the bit processing table before dividing or what is divided from that element, by adding it to other element or to what is divided from the other elements. The adjusted exposure table generated as described above is stored in the storage device 144. The adjusted exposure table should be generated before starting writing processing. Although, in the example described above, the adjusted exposure table is generated in the writing apparatus 100, it is not limited thereto. If the digit number “n” which is used when converting the irradiation time per shot into binary number data has been previously set, the adjusted exposure table itself can also be set beforehand. Therefore, it is also preferable to generate an adjusted exposure table outside the apparatus in advance, and to input it to the writing apparatus 100 to be stored in the storage device 144. In the irradiation time arrangement data processing step (S109), referring to the bit processing table stored in the storage device 144, the bit processing unit 70 converts n-digit binary number data, which was converted in the binary digit conversion step (S108), into (n+a)-digit binary number data. For example, in the case of the bit processing table of FIG. 11, 10-digit binary number data is converted to 14-digit binary number data. For example, if N=5Δ, ten digits “0000110010” is converted to fourteen digits “00000000110010”. For example, if N=500, similarly, ten digits “0111110100” is converted to fourteen digits “00001111110100”. In that case, since the ninth digit of the ten-digit binary number data is “1” and the tenth digit of it is “0”, the ninth digit and the tenth digit, (8a and 8b), of the fourteen digits after processing are “1”, and the eleventh digit to the fourteenth digit, (9a, 9b, 9c, and 9d), are “0”. For example, if N=700, similarly, ten digits “1010111100” is converted to fourteen digits “11110010111100”. For example, if N=1023, similarly, ten digits “1111111111” is converted to fourteen digits “11111111111111”. In the irradiation time arrangement data output step (S110), the transmission processing unit 68 outputs, for each beam shot, irradiation time arrangement data having been converted to (n+a)-digit binary number data, to the deflection control circuit 130. In that case, referring to the grouped exposure table stored in the storage device 144, the transmission processing unit 68 outputs, for each group, irradiation time arrangement data to the deflection control circuit 130. In the target group data transmission step (S112), the deflection control circuit 130 outputs, for each shot, irradiation time arrangement data of each group to the logic circuit 41 for each beam. Moreover, synchronized with this, the deflection control circuit 130 outputs timing data of each irradiation step to the logic circuit 132 for common blanking. Since the shift register 40 is used for the logic circuit 41 as shown in FIG. 5 in the first embodiment, the deflection control circuit 130 transmits data of each bit (the same digit number) configuring the same group to each logic circuit 41 of the blanking plate 204 in the order of beam arrangement (or in the order of identification number). Moreover, a clock signal (CLK1) for synchronization, a read signal (read) for data read-out, and an adder signal (BLK) are output. For example, as the data of the k1-th bit (the k1-th digit) and the k2-th bit (the k2-th digit) that configure the k-th group of the beam 1, two bits “11” are generated. As the data of the k1-th bit (the k1-th digit) and the k2-th bit (the k2-th digit) that configure the k-th group of the beam 2, two bits “11” are generated. As the data of the k1-th bit (the k1-th digit) and the k2-th bit (the k2-th digit) that configure the k-th group of the beam 3, two bits “00” are generated. As the data of the k1-th bit (the k1-th digit) and the k2-th bit (the k2-th digit) that configure the k-th group of the beam 4, two bits “11” are generated. As the data of the k1-th bit (the k1-th digit) and the k2-th bit (the k2-th digit) that configure the k-th group of the beam 5, two bits “00” are generated. From the end beam side, the deflection control circuit 130 transmits each 2-bit data of “0011001111”. Then, from the upper side shift register to the next one, the shift register 40 of each beam transmits data in order, two bits by two bits, according to a clock signal (CLK1). For example, with respect to the data of the k-th group of the beams 1 to 5, by five clock signals, 2-bit data “11” is stored in the shift register 40 of the beam 1, 2-bit data “11” is stored in the shift register 40 of the beam 2, 2-bit data “00” is stored in the shift register 40 of the beam 3, 2-bit data “11” is stored in the shift register 40 of the beam 4, and 2-bit data “00” is stored in the shift register 40 of the beam 5. Next, when inputting a read signal (read), the register 42 of each beam reads in the k-th group data of each beam from the shift register 40. In the example described above, as the data of the k-th group, 2-bit data “11” is stored in the register 42 of the beam 1, 2-bit data “11” is stored in the register 42 of the beam 2, 2-bit data “00” is stored in the register 42 of the beam 3, 2-bit data “11” is stored in the register 42 of the beam 4, and 2-bit data “00” is stored in the register 42 of the beam 5. When inputting the data of the k-th group, the individual register 42 of each beam outputs, according to the data, an ON/OFF signal to the AND computing unit 44 through the selector 48. In the first embodiment, the output of the individual register 42 is switched from the output of the k1-th bit (the k1-th digit) to the output of the k2-th bit (the k2-th digit) by switching the selector 48. When the selector 48 inputs a select signal (select), one is switched to the other in the 2-bit signal. If the BLK signal is an ON signal and the signal of the register 42 is ON, the AND computing unit 44 outputs an ON signal to the amplifier 46, and then, the amplifier 46 applies an ON voltage to the electrode 24 of the individual blanking deflector. In other case, the AND computing unit 44 outputs an OFF signal to amplifier 46, and then, the amplifier 46 applies an OFF voltage to the electrode 24 of the individual blanking deflector. While the 2-bit data of the k-th group is being processed, the deflection control circuit 130 transmits the data of the next (k+1)th group to each logic circuit 41 of the blanking plate 204, in the order of beam arrangement (or in the order of identification number). Hereafter, it should similarly proceed to the data processing of the last group. The AND computing unit 44 shown in FIG. 5 may be omitted. However, it is effective in that a beam can be controlled to be OFF by the AND computing unit 44 in the case of not being able to make the beam OFF because of a trouble of one of elements of the logic circuit 41. In the writing step (S114) according to irradiation time of a target group, writing is performed, for each beam shot, based on irradiation time of each irradiation step of a target group, in irradiation divided into a plurality of irradiation steps of a plurality of groups. FIG. 14 is a timing chart showing a beam ON/OFF switching operation with respect to a part of an irradiation step in one shot according to the first embodiment. FIG. 14 shows one beam (beam 1) in a plurality of beams that configure a multi-beam. Here, for example, there are shown irradiation steps from the k group composed of the k1-th bit (the k1-th digit) and the k2-th bit (the k2-th digit) of the beam 1 to the (k+1) group composed of the k3-th bit (the k3-th digit) and the k4-th bit (the k4-th digit). The irradiation time arrangement data shows the case of, for example, the k1-th bit (the k1-th digit) being “1”, the k2-th bit (the k2-th digit) being “1”, the k3-th bit (the k3-th digit) being “0”, and the k4-th bit (the k4-th digit) being “1”. First, in response to an input of a read signal of the k group composed of the k1-th bit (the k1-th digit) and the k2-th bit (the k2-th digit), the individual register 42 (an individual register signal 1 and an individual register signal 2) outputs an ON/OFF signal, according to the stored data (two bits) of the k1-th bit (the k1-th digit) and the k2-th bit (the k2-th digit). In the first embodiment, since a 2-bit signal is used, it is necessary to perform selecting and switching the signal. In FIG. 14, first, data of the individual register 1 is selected by the selector 48, and an ON signal of the k1-th bit (the k1-th digit) is output to the individual amplifier. Next, with respect to an output of the individual register 42, data of the individual register 2 is selected by the switching of the selector 48, and the output of the k1-th bit (the k1-th digit) is switched to the output of the k2-th bit (the k2-th digit). Hereafter, switching like this is serially repeated for each irradiation step. Since the data of the k1-th bit (k1-th digit) is ON data, the individual amplifier 46 (an individual amplifier 1) outputs an ON voltage to be applied to the blanking electrode 24 for the beam 1. On the other hand, in the logic circuit 132 for common blanking, ON or OFF is switched depending upon timing data of each irradiation step of (n+a) bits (e.g., ten bits). In the common blanking system, an ON signal is output during the irradiation time of each irradiation step of each group. For example, if Δ=1 ns, the irradiation time of the first irradiation step (e.g., the irradiation step of k=0) is Δ×1=1 ns. The irradiation time of the second irradiation step (e.g., the irradiation step of k=9d (the fourteenth digit)) is Δ×128=128 ns. The irradiation time of the first irradiation step (e.g., the irradiation step of k=1) of the group 2 is Δ×2=2 ns. The irradiation time of the second irradiation step (e.g., the irradiation step of k=9c (the thirteenth digit)) is Δ×128=128 ns. Similarly, it becomes ON during the irradiation time of each irradiation step of each group, hereinafter. In the logic circuit 132, when inputting timing data of each irradiation step, the register 50 outputs ON data of the k-th digit (the k-th bit), the counter 52 counts the irradiation time of the k-th digit (the k-th bit), and controlling is performed to be OFF after the irradiation time has passed. Hereafter, beam irradiation is performed for each group in order. As described above, according to the first embodiment, data transmission time period can be included in the total of the grouped irradiation time period in the irradiation step. In the common blanking system, compared with ON/OFF switching of the individual blanking system, ON/OFF switching is performed after the voltage stabilization time (settling time) S1/S2 of the amplifier 46 has passed. In the example of FIG. 14, after the individual amplifier 1 has become ON and the settling time S1 of the individual amplifier 1 at the time of switching from OFF to ON has passed, the common amplifier becomes ON. Thereby, beam irradiation at an unstable voltage at the time of rise of the individual amplifier 1 can be eliminated. Then, the common amplifier becomes OFF when the irradiation time of the k-th digit (the k-th bit) being a target has passed. Consequently, in the case of both the individual amplifier and the common amplifier being ON, an actual beam becomes ON, and irradiates the target object 101. Therefore, it is controlled such that the ON time period of the common amplifier becomes an actual beam irradiation time period. On the other hand, in the case where the common amplifier becomes ON when the individual amplifier 1 is OFF, after the individual amplifier 1 becomes OFF and the settling time S2 of the individual amplifier 1 at the time of switching from ON to OFF has passed, the common amplifier becomes ON. Thereby, beam irradiation at an unstable voltage at the time of fall of the individual amplifier 1 can be eliminated. As described above, in the individual beam ON/OFF switching step (S116), beam ON/OFF control is individually performed for each corresponding beam in multiple beams by a plurality of individual blanking systems (blanking plate 204, etc.), and, for each beam, with respect to each irradiation step (irradiation) of the k-th group, beam ON/OFF switching is performed by the individual blanking system for the beam concerned. In the example of FIG. 14, since the irradiation step of the k2-th digit (the k2-th bit) of the k-th group is not beam OFF, switching from ON to OFF is not performed. However, for example, if the irradiation step of the k2-th digit (the k2-th bit) is beam OFF, it should be understood that switching from ON to OFF is performed. In the common beam ON/OFF switching step (S118), for each beam, with respect to each irradiation step (irradiation) of the k-th group, after performing beam ON/OFF switching by the individual blanking system, beam ON/OFF controlling is performed all at once for the whole of the multiple beams by using the common blanking system (the logic circuit 132, the deflector 212, etc.), and blanking control is performed so that it may be in a beam ON state during the irradiation time corresponding to each irradiation step (irradiation) of the k-th group. As described above, since there is a restriction on the installation area of the circuit and the current to be used in the circuit in the blanking plate 204, a simple amplifier circuit is used. Therefore, it is also limited in reducing the settling time of the individual amplifier. By contrast, in the common blanking system, a highly precise amplifier circuit of sufficient size, current, and scale can be installed outside the optical column. Therefore, the settling time of the common amplifier can be shortened. Thus, according to the first embodiment, by making beam ON by the common blanking system after becoming beam ON by the individual blanking system (or after a read signal of a target digit is output) and after the settling time having passed, it becomes possible to eliminate a voltage unstable time of the individual amplifier and a noise component containing crosstalk, on the blanking plate, and to perform a blanking operation based on a highly precise irradiation time. In the determination step (S120), the writing control unit 72 determines, with respect to irradiation time arrangement data, whether transmission of data of all the groups has been completed or not. When it has not been completed yet, it proceeds to the group change step (S122). When it has been completed, it proceeds to the determination step (S124). In the group change step (S122), the writing control unit 72 changes a target group. For example, the writing control unit 72 changes the target group (or “target digit”) from the k-th group to the (k+1)th group. Then, it returns to the data transmission step (S112) of the target group. With respect to the processing of the (k+1)th group, steps from the data transmission step (S112) to the group change step (S122) of the target group are executed. Then, it is similarly repeated until data processing of irradiation time arrangement data of all the groups has been completed in the determination step (S120). As described above, a maximum irradiation time period (2n−1) per shot of beams of the multiple beams is divided into a plurality, being a digit number “n” set in advance, of irradiation time periods (first irradiation time periods), each of which is calculated by multiplying a corresponding gray scale value (second gray scale value) of a plurality of gray scale values (second gray scale values) by a quantization unit Δ, where the plurality of gray scale values (second gray scale values) are gray scale values defined in decimal numbers converted from each digit value of data of binary numbers of the digit number “n”. A plurality of irradiation time periods irradiation time periods (second irradiation time periods), which are a part of the plurality of irradiation time periods (first irradiation time periods), are divided into a plurality of irradiation time periods (third irradiation time periods). Further, irradiation of each beam of the multiple charged particle beams by using the plurality of irradiation time periods (third irradiation time periods) and remaining undivided plurality of irradiation time periods (first irradiation time periods), is divided into irradiation steps (first irradiation steps) of the plurality of irradiation time periods (third irradiation time periods) and irradiation steps (second irradiation steps) of the remaining undivided plurality of irradiation time periods (first irradiation time periods). Then, a target object is irradiated, in order, with the multiple beams such that the plurality of groups are respectively composed of combination of at least two irradiation steps of the irradiation steps (first irradiation steps) and the irradiation steps (second irradiation steps) and the plurality of groups continue in order, for each group of a plurality of groups in each shot of beams. The electron beam 200 emitted from the electron gun assembly 201 (emission unit) almost perpendicularly illuminates the whole of the aperture member 203 by the illumination lens 202. A plurality of holes (openings), each being a quadrangle, are formed in the aperture member 203. The region including all the plurality of holes is irradiated with the electron beam 200. For example, a plurality of quadrangular electron beams (multiple beams) 20a to 20e are formed by letting parts of the electron beam 200 irradiating the positions of a plurality of holes pass through a corresponding hole of the plurality of holes of the aperture member 203 respectively. The multiple beams 20a to 20e respectively pass through a corresponding blanker (the first deflector: individual blanking system) of the blanking plate 204. Each blanker respectively deflects (performs blanking deflection) the electron beam 20 passing individually. FIG. 15 is a schematic diagram explaining a blanking operation according to the first embodiment. The multiple beams 20a, 20b, . . . , 20e, having passed through the blanking plate 204 are reduced by the reducing lens 205, and go toward the hole at the center of the limiting aperture member 206. At this stage, the electron beam 20 which was deflected by the blanker of the blanking plate 204 deviates from the hole of the center of the limiting aperture member 206 (blanking aperture member) and is blocked by the limiting aperture member 206. On the other hand, if the electron beam 20 which was not deflected by the blanker of the blanking plate 204 is not deflected by the deflector 212 (common blanking system), it passes through the hole at the center of the limiting aperture member 206, as shown in FIG. 1. Blanking control is performed by combination of ON/OFF of the individual blanking system and ON/OFF of the common blanking system so as to control ON/OFF of the beam. Thus, the limiting aperture member 206 blocks each beam which was deflected to be a beam OFF state by the individual blanking system or the common blanking system. Then, beam of an irradiation step obtained by dividing one beam shot is formed by beams having been made during from a beam ON state to a beam OFF state and having passed through the limiting aperture member 206. The multi-beams 20 having passed through the limiting aperture member 206 are focused by the objective lens 207 in order to be a pattern image of a desired reduction ratio, and respective beams (the entire multi-beams 20) having passed through the limiting aperture member 206 are collectively deflected in the same direction by the deflector 208 so as to irradiate respective irradiation positions on the target object 101. While the XY stage 105 is continuously moving, controlling is performed by the deflector 208 so that irradiation positions of beams may follow the movement of the XY stage 105, for example. Ideally, multi-beams 20 to irradiate at a time are aligned at pitches obtained by multiplying the arrangement pitch of a plurality of holes of the aperture member 203 by a desired reduction ratio described above. The writing apparatus 100 performs a writing operation by the raster scan method which continuously irradiates shot beams in order, and when writing a desired pattern, a required beam is controlled by blanking control to be ON according to the pattern. In the determination step (S124), the writing control unit 72 determines whether all the shots have been completed. If all the shots have been completed, it ends. If all the shots have not been completed yet, it returns to the gray level value N calculation step (S106), and the steps from the gray level value N calculation step (S106) to the determination step (S124) are repeated until all the shots have been completed. FIG. 16 is a conceptual diagram explaining a writing operation according to the first embodiment. As shown in FIG. 16, a writing region 30 of the target object 101 is virtually divided into a plurality of strip-shaped stripe regions 32 each having a predetermined width in the y direction, for example. Each of the stripe regions 32 serves as a writing unit region. The XY stage 105 is moved and adjusted such that an irradiation region 34 to be irradiated with one-time irradiation of the multi-beams 20 is located at the left end of the first stripe region 32 or at a position more left than the left end, and then writing is started. When writing the first stripe region 32, the writing advances relatively in the x direction by moving the XY stage 105 in the −x direction, for example. The XY stage 105 is continuously moved at a predetermined speed, for example. After writing the first stripe region 32, the stage position is moved in the −y direction and adjusted such that the irradiation region 34 is located at the right end of the second stripe region 32 or at a position more right than the right end and located to be relatively in the y direction. Then, similarly, writing advances in the −x direction by moving the XY stage 105 in the x direction, for example. That is, writing is performed while alternately changing the direction, such as performing writing in the x direction in the third stripe region 32, and in the −x direction in the fourth stripe region 32, and thus, the writing time can be reduced. However, the writing operation is not limited to the case of performing writing while alternately changing the direction, and it is also acceptable to perform writing in the same direction when writing each stripe region 32. By one shot, a plurality of shot patterns of the same number as the holes 22 are formed at a time by multiple beams which have been formed by passing through respective corresponding holes 22 of the aperture member 203. FIGS. 17A to 17C are conceptual diagrams explaining examples of a writing operation in a stripe according to the first embodiment. The examples of FIGS. 17A to 17C show the cases where writing is performed in a stripe by using multiple beams of 4×4 in the x and y directions, for example. The examples of FIGS. 17A to 17C show the cases where a stripe region is divided in the y direction by twice the width of an irradiation region of the whole multi-beam, for example. There is shown the case where exposure (writing) of one irradiation region of the whole of multiple beams is completed by four shots (one shot is a total of a plurality of irradiation steps) performed while shifting the irradiation position by one mesh in the x direction or the y direction. First, the upper region of the stripe region is to be written. FIG. 17A shows the mesh region irradiated by the first one-shot (one shot is a total of a plurality of irradiation steps). Next, as shown in FIG. 17B, the second one-shot (one shot is a total of a plurality of irradiation steps) is performed while shifting the position in the y direction to the mesh region not having been irradiated yet. Next, as shown in FIG. 17C, the third one-shot (one shot is a total of a plurality of irradiation steps) is performed while shifting the position in the x direction to the mesh region not having been irradiated yet. FIGS. 18A to 18C are conceptual diagrams explaining examples of a writing operation in a stripe according to the first embodiment. FIGS. 18A to 18C are continued from FIG. 17C. As shown in FIG. 18A, the fourth one-shot (one shot is a total of a plurality of irradiation steps) is performed while shifting the position in the y direction to the mesh region not having been irradiated yet. Exposure (writing) of one irradiation region of the whole of multiple beams is completed by these four shots (one shot is a total of a plurality of irradiation steps). Next, the lower region of the stripe region is to be written. As shown in FIG. 18B, the lower region of the stripe region is irradiated by the first one-shot (one shot is a total of a plurality of irradiation steps). Next, the second one-shot (one shot is a total of a plurality of irradiation steps) is performed while shifting the position in the y direction to the mesh region not having been irradiated yet. Next, the third one-shot (one shot is a total of a plurality of irradiation steps) is performed while shifting the position in the x direction to the mesh region not having been irradiated yet. The fourth one-shot (one shot is a total of a plurality of irradiation steps) is performed while shifting the position in the y direction to the mesh region not having been irradiated yet. By the operations described above, writing of the first row of the irradiation region of multiple beams in the stripe region is completed. Then, as shown in FIG. 18C, writing is to be similarly performed for the second row of the multiple beam irradiation region while shifting the position in the x direction. The whole stripe region can be written by repeating the operations described above. FIGS. 19A to 19C are conceptual diagrams explaining other examples of a writing operation in a stripe according to the first embodiment. FIGS. 19A to 19C show examples in which writing in a stripe is performed using 4×4 multiple beams in the x and y directions. The examples of FIG. 19A to FIG. 19C show the case where there is a distance between beams and a stripe region is divided in the y direction by a width somewhat greater than or equal to the irradiation region of the whole of multiple beams, for example. Exposure (writing) of one irradiation region by the whole of multiple beams is completed by sixteen shots (one shot is a total of a plurality of irradiation steps) performed while shifting the irradiation position by one mesh in the x direction or the y direction. FIG. 19A shows the mesh region irradiated by the first one-shot (one shot is a total of a plurality of irradiation steps). Next, as shown in FIG. 19B, the second one-shot, the third one-shot, and the fourth one-shot (one shot is a total of a plurality of irradiation steps) are performed while shifting the position by one mesh, one by one, in the y direction to the mesh region not having been irradiated yet. Next, as shown in FIG. 19C, the fifth one-shot (one shot is a total of a plurality of irradiation steps) is performed while shifting the position by one mesh in the x direction to the mesh region not having been irradiated yet. Next, the sixth one-shot, the seventh one-shot, and the eighth one-shot (one shot is a total of a plurality of irradiation steps) are performed while shifting the position by one mesh, one by one, in the y direction to the mesh region not having been irradiated yet. FIGS. 20A to 20C are conceptual diagrams explaining other examples of a writing operation in a stripe according to the first embodiment. FIGS. 20A to 20C are continued from FIG. 19C. As shown in FIG. 20A, the ninth one-shot to the sixteenth one-shot (one shot is a total of a plurality of irradiation steps) are repeatedly performed in order similarly to the operations of FIGS. 19A to 19C. The examples of FIGS. 19A to 19C and 20A to 20C show the case of performing multi-pass writing (multiplicity=2), for example. In such a case, the irradiation position is shifted in the x direction by about half the size of the irradiation region of the whole of multiple beams, and as shown in FIG. 20B, the first one-shot (one shot is a total of a plurality of irradiation steps) of the second layer of the multi-pass writing is performed. As described referring to FIGS. 19B and 19C, the second one-shot to the eighth one-shot (one shot is a total of a plurality of irradiation steps) of the second layer of the multi-pass writing are performed one by one, hereinafter. As shown in FIG. 20C, the ninth one-shot to the sixteenth one-shot (one shot is a total of a plurality of irradiation steps) are to be repeatedly performed in order similarly to the operations of FIGS. 19B to 19C. As described above, according to the first embodiment, the latency time of beam irradiation operation due to data transmission time can be reduced or avoided while maintaining the restriction on a circuit installation space. Moreover, according to the first embodiment, the precision of irradiation time control and, further, the precision of dose control can be improved while maintaining the restriction on a circuit installation space. Furthermore, since the data amount of the logic circuit 41 of the individual blanking system is one bit, power consumption can be suppressed. Although the first embodiment shows the case where the quantization unit Δ (a counter period of the common blanking system) is set uniquely, it is not limited thereto. The second embodiment describes the case where the quantization unit Δ is set variably. The apparatus structure according to the second embodiment is the same as that of FIG. 1. The flowchart showing main steps of a writing method according to the second embodiment is the same as that of FIG. 6. The content of the second embodiment is the same as that of the first embodiment except what is particularly described below. FIGS. 21A to 21E are time charts for comparing the exposure latency time according to the second embodiment. FIG. 21A shows an example of performing beam irradiation or not performing beam irradiation of each beam in each irradiation step when dividing one shot into “n” irradiation steps. In the case of dividing a shot into “n” irradiation steps, the irradiation time per shot is (2n−1) at the maximum. FIG. 21A shows the case of n=10, as an example. In such a case, the irradiation time per shot is 1023Δ at the maximum. In FIG. 21A, the irradiation time per shot is divided into irradiation steps of 10 times: 512 Δ, 256Δ, 128Δ, 64Δ, 32Δ, 16Δ, 8Δ, 4Δ, 2Δ, and 1Δ, which are described in order from the longer irradiation time. In FIG. 21A, irradiation steps whose irradiation time is less than 128Δ are not shown. In FIG. 21A, the beam 1 is OFF (no beam irradiation) in the irradiation step whose irradiation time is 128Δ, ON (beam irradiation) in the irradiation step whose irradiation time is 256Δ, and ON (beam irradiation) in the irradiation step whose irradiation time is 512Δ. The beam 2 is ON (beam irradiation) in the irradiation step whose irradiation time is 128Δ, ON (beam irradiation) in the irradiation step whose irradiation time is 256Δ, and OFF (no beam irradiation) in the irradiation step whose irradiation time is 512Δ. The beam 3 is OFF (no beam irradiation) in the irradiation step whose irradiation time is 128Δ, ON (beam irradiation) in the irradiation step whose irradiation time is 256Δ, and OFF (no beam irradiation) in the irradiation step whose irradiation time is 512Δ. The beam 4 is ON (beam irradiation) in the irradiation step whose irradiation time is 128Δ, ON (beam irradiation) in the irradiation step whose irradiation time is 256Δ, and OFF (no beam irradiation) in the irradiation step whose irradiation time is 512Δ. The beam 5 is OFF (no beam irradiation) in the irradiation step whose irradiation time is 128Δ, ON (beam irradiation) in the irradiation step whose irradiation time is 256Δ, and OFF (no beam irradiation) in the irradiation step whose irradiation time is 512Δ. FIG. 21B shows an example of a total irradiation time per shot of each beam shown in FIG. 21A. FIG. 21B shows, as a comparative example, the case where the quantization unit Δ is set uniquely. Moreover, with respect to each beam shown in FIG. 21A, irradiation steps whose irradiation time is less than 128Δ are OFF (no beam irradiation). In such a case, as shown in FIG. 21B, the total irradiation time per shot of the beam 1 is 768Δ, for example. The total irradiation time per shot of the beam 2 is 384Δ, for example. The total irradiation time per shot of the beam 3 is 256Δ, for example. The total irradiation time per shot of the beam 4 is 384Δ, for example. The total irradiation time per shot of the beam 5 is 256Δ, for example. On the other hand, as described above, the irradiation time per shot is 1023Δ at the maximum. When the total irradiation time per shot of beams of each beam is shorter than the maximum irradiation time, a latency time occurs as shown in FIG. 21B. Then, in the second embodiment, the quantization unit Δ is made to be variable in order to reduce such a latency time. As shown in FIG. 21C, the quantization unit Δ is set such that the maximum value of the irradiation time per shot of beams corresponds to the total irradiation time per shot (a sum of irradiation time of irradiation steps per shot) of a beam in the case where the total irradiation time per shot of all the beams of multiple beams of all the shots is the maximum. In the example of FIG. 21B, the total irradiation time per shot of the beam 1 is 768Δ, and is the maximum. Therefore, a quantization unit Δ1 is set such that the maximum irradiation time 768Δ per shot corresponds to 1023Δ1. Thereby, the repetition period (interval) of each shot can be shortened. FIG. 21D shows, treating the maximum irradiation time 768Δ as 1023Δ1, an example of irradiation or no irradiation of each beam in each irradiation step in the case of again dividing one shot into ten irradiation steps. In FIG. 21D, irradiation steps whose irradiation time is less than 128Δ are not shown. Since the beam 1 in FIG. 21D is a beam being a standard of a repetition period, it is set to be in the ON state (beam irradiation) in all the irradiation steps. Since the beams 2 and 4 are 384Δ, when converted, they become about 512Δ1. Therefore, they are ON (beam irradiation) in the irradiation step whose irradiation time is 512Δ1, and OFF (no beam irradiation) in the other irradiation steps. Since beams 3 and 5 are 256Δ, when converted, they become 341Δ1. Therefore, they are ON (beam irradiation) in the irradiation steps whose irradiation time is 256Δ1, 64Δ1, 16Δ1, 4Δ1, or 1Δ1, and OFF (no beam irradiation) in the other irradiation steps. In FIG. 21E, for each shot, the quantization unit Δ is set such that the maximum value of the irradiation time per shot corresponds to the total irradiation time per shot of a beam in the case where the total irradiation time per shot of all the beams of multiple beams is the maximum. In the example of FIG. 21E, the total irradiation time per shot of the first one-shot of the beam 1 is 768Δ, which is the maximum. Therefore, the quantization unit Δ1 is set such that the maximum irradiation time 768Δ per shot corresponds to 1023Δ1. Thereby, the repetition period (interval) of the first one-shot can be shortened. Moreover, the total irradiation time per shot of the second one-shot of the beam 2 is 640Δ, which is the maximum. Therefore, the quantization unit Δ2 is set such that the maximum irradiation time 640Δ per shot corresponds to 1023Δ2. Thereby, the repetition period (interval) of the second one-shot can be shortened. Similarly, for each shot, Δ3, Δ4, . . . is to be set, hereinafter. As described above, the quantization unit Δ is made to be variable. Thereby, the latency time can be suppressed. Therefore, writing time can be shortened. Although the case of n=10 is shown as an example in FIGS. 21A to 21E, other case, namely the case other than n=10, is also similarly applicable. As described above, according to the second embodiment, it is possible to reduce or suppress the latency time at the time of performing irradiation steps. Although, in each embodiment described above, blanking control is performed for each of a plurality of irradiation steps made by dividing one shot, for each beam, by using the blanking plate 204 for individual blanking control and the deflector 212 for common blanking, it is not limited thereto. In the third embodiment, there will be described a configuration in which blanking control is performed for each of a plurality of irradiation steps made by dividing one shot, for each beam, by using the blanking plate 204 for individual blanking control without using the deflector 212 for common blanking. FIG. 22 is a schematic diagram showing the structure of a writing apparatus according to the third embodiment. FIG. 22 is the same as FIG. 1 except that the deflector 212 does not exist and output of the logic circuit 132 is connected to the blanking plate 204. Main steps of a writing method according to the third embodiment are the same as those of FIG. 6. The content of the third embodiment is the same as that of the first embodiment except what is particularly described below. FIG. 23 is a schematic diagram showing the internal structure of an individual blanking control circuit and a common blanking control circuit according to the third embodiment. The content of FIG. 23 is the same as that of FIG. 5 except that the deflector 212 does not exist and an output signal of the logic circuit 132 is input into the AND computing unit 44 (AND circuit) instead of a signal from the deflection control circuit 130. In the individual beam ON/OFF switching step (S116), an ON/OFF control signal (first ON/OFF control signal) for a beam is output by the logic circuit (first logic circuit) of the beam concerned, for each beam, with respect to each of a plurality of times of irradiation, by using a plurality of logic circuits (first logic circuit) each including the shift register 40 and the individual register 42 each respectively outputting a beam ON/OFF control signal to a corresponding beam in multiple beams. Specifically, as described above, when inputting 2-bit data of the k-th group, the individual register 42 of each beam outputs an ON/OFF signal to the AND computing unit 44 through the selector 48 based on the input data. If the data of the k-th group is “11”, two ON signals are to be output, and if the data is “00”, two OFF signals are to be output. In the common beam ON/OFF switching step (S118), for each beam, with respect to each of a plurality of times of irradiation, after a beam ON/OFF control signal has been switched by the logic circuit for individual blanking, a beam ON/OFF control signal (second ON/OFF control signal) is output so that a beam may be in the ON state during the irradiation time corresponding to the irradiation concerned, by using the logic circuit 132 (second logic circuit) which collectively outputs a beam ON/OFF control signal to the whole of multiple beams. Specifically, in the logic circuit 132 for common blanking, ON/OFF is switched depending upon 10-bit timing data of each irradiation step. The logic circuit 132 outputs an ON/OFF control signal to the AND computing unit 44. In the logic circuit 132, an ON signal is output during the irradiation time of each irradiation step. In the blanking control step, the AND computing unit 44 performs blanking control so that a beam concerned may be in the ON state during the irradiation time corresponding to the irradiation concerned, when both the ON/OFF control signal for an individual beam and the ON/OFF control signal for a common beam are ON control signals. When both the ON/OFF control signal for an individual beam and the ON/OFF control signal for a common beam are ON control signals, the AND computing units 44 outputs an ON signal to the amplifier 46, and, then, the amplifier 46 applies an ON voltage to the electrode 24 of the individual blanking deflector. In other case, the AND computing unit 44 outputs an OFF signal to the amplifier 46, and, then, the amplifier 46 applies an OFF voltage to the electrode 24 of the individual blanking deflector. Thus, when both the ON/OFF control signal for an individual beam and the ON/OFF control signal for a common beam are ON control signals, the electrode 24 (an individual blanking system) of the individual blanking deflector individually performs beam ON/OFF control so that the beam concerned may be in the ON state during the irradiation time corresponding to the irradiation concerned. As described above, even when the blanking plate 204 for individual blanking control is used without using the deflector 212 for common blanking, the restriction on a circuit installation space can be maintained similarly to the first embodiment. Moreover, since the logic circuit 41 for individual blanking has a data amount of one bit, power consumption can also be suppressed. Furthermore, there is an advantage that the deflector 212 for common blanking can be omitted. In each embodiment described above, each logic circuit 41 for individual blanking control is arranged on the blanking plate 204, but, however, it may be arranged outside. In the fourth embodiment, the case of arranging each logic circuit 41 for individual blanking control outside the blanking plate 204 will be described. The apparatus structure according to the fourth embodiment is the same as that of FIG. 1 except that each logic circuit 41 for individual blanking control is arranged at the outside of the blanking plate 204. The flowchart showing main steps of a writing method according to the fourth embodiment is the same as that of FIG. 6. The content of the fourth embodiment is the same as that of one of the first to third embodiments except what is particularly described below. FIG. 24 is a schematic diagram explaining the arrangement state between the logic circuit and the blanking plate 204 according to the fourth embodiment. In the fourth embodiment, each logic circuit 41 for individual blanking control and each amplifier 46 are arranged in the logic circuit 134 arranged outside the writing unit 150, and connected to each electrode 24 for individual blanking control by wiring. In such a structure, since the wiring becomes long, crosstalk and settling time increase. However, as described above, according to the fourth embodiment, since ON/OFF switching is performed by the common blanking system after having performed ON/OFF switching by the individual blanking system and having waited for voltage stability, the irradiation time can be controlled highly accurately without being affected by crosstalk and settling time even if they increase. In the above embodiments, the case where each group is composed of two irradiation steps has been described, but however, it is not limited thereto. In the fifth embodiment, the case where each group is composed of three or more irradiation steps will be described. Hereafter, the contents of the fifth embodiment are the same as those of one of the embodiments described above except what is particularly explained below. FIG. 25 shows an example of a grouped exposure table according to the fifth embodiment. FIG. 25 shows the case where irradiation per shot is divided into irradiation steps of twelve irradiation time periods (exposure time), for example. The example of FIG. 25 has a configuration so that the difference between the totals of grouped irradiation time composed of three irradiation steps may become smaller to be close to uniform. As the exposure step 1, the group 1 is composed of the irradiation time Δ, the irradiation time 128Δ, and the irradiation time 128Δ. As the exposure step 2, the group 2 is composed of the irradiation time 32Δ, the irradiation time 64Δ, and the irradiation time 158Δ. As the exposure step 3, the group 3 is composed of the irradiation time 8Δ, the irradiation time 2Δ, and the irradiation time 256Δ. As the exposure step 4, the group 4 is composed of the irradiation time 4Δ, the irradiation time 16Δ, and the irradiation time 226Δ. 128Δ, 158Δ, and 226Δ being the exposure time 3 of the exposure steps 1, 2, and 4 are obtained by dividing 512Δ. That is, it is also preferable to make groups each composed of three irradiation steps as described above. In such a case, the data of each group is 3-bit data. FIG. 26 shows another example of a grouped exposure table according to the fifth embodiment. FIG. 26 shows the case where irradiation per shot is divided into irradiation steps of twelve irradiation time periods (exposure time), for example. The example of FIG. 26 has a configuration so that the difference between the totals of grouped irradiation time composed of four irradiation steps may become smaller to be close to uniform. As the exposure step 1, the group 1 is composed of the irradiation time 4Δ, the irradiation time 16Δ, the irradiation time 64Δ, and the irradiation time 256Δ. As the exposure step 2, the group 2 is composed of the irradiation time 2Δ, the irradiation time 8Δ, the irradiation time 128Δ, and the irradiation time 204Δ. As the exposure step 3, the group 3 is composed of the irradiation time 1Δ, the irradiation time 32Δ, the irradiation time 128Δ, and the irradiation time 180Δ. 204Δ being the exposure time 4 of the exposure step 2, and 128Δ and 180Δ being the exposure time 3 and the exposure time 4 of the exposure step 3 are obtained by dividing 512Δ. That is, it is also preferable to make groups each composed of four irradiation steps as described above. In such a case, the data of each group is 4-bit data. As has been described above, it is also preferable to configure each group by three or more irradiation steps. Embodiments have been explained referring to concrete examples described above. However, the present invention is not limited to these specific examples. While the apparatus configuration, control method, and the like not directly necessary for explaining the present invention are not described, some or all of them may be suitably selected and used when needed. For example, although description of the configuration of a control unit for controlling the writing apparatus 100 is omitted, it should be understood that some or all of the configuration of the control unit is to be selected and used appropriately when necessary. In addition, any other multi charged particle beam writing apparatus and multi charged particle beam writing method that include elements of the present invention and that can be appropriately modified by those skilled in the art are included within the scope of the present invention. Additional advantages and modification will readily occur to those skilled in the art. Therefore, the invention in its broader aspects is not limited to the specific details and representative embodiments shown and described herein. Accordingly, various modifications may be made without departing from the spirit or scope of the general inventive concept as defined by the appended claims and their equivalents. |
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abstract | Provided are systems and methods for multi-channel non-destructive inspection which provide high data throughput, logarithmic amplification of large dynamic range, and simplicity of supporting electronics. More specifically, provided are systems and methods for inspecting a structure that may use an interface board, two pulser boards, each coupled to 16 transmit channels, and two receiver boards, each coupled to 16 receive channels, where the receiver boards are capable of processing data from the 32 receive channels by logarithmically amplifying at least 70 dB of dynamic range. A receiver board may include a serial connection of two layers of multiplexing switches to provide 70 dB isolation between channels, a logarithmic amplifier for logarithmically amplifying 70 dB of dynamic range, a linear amplifier, and an analog-to-digital converter. |
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abstract | A nuclear power plant comprises a pressurized water reactor (PWR) and a steam generator driving a turbine driving an electric generator. A condenser condenses steam after flowing through the turbine. Responsive to a station blackout, the nuclear power plant is electrically isolated and a bypass valve is opened to convey bypass steam flow from the steam generator to the condenser without flowing through the turbine. The thermal power output of the PWR is gradually reduced over the transition time interval. After opening, the bypass valve is gradually closed over the transition time interval. A supplemental bypass valve may also be opened responsive to the station blackout to convey supplemental bypass steam flow from the steam generator to a feedwater system supplying secondary coolant feedwater to the steam generator. The supplemental bypass steam flow does not flow through the turbine and does not flow through the condenser. |
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claims | 1. A collimator having an adjustable focal length for an X-ray inspection system, the collimator comprising:an outer part having a conical inner surface;an inner part having a conical outer surface, the outer part and the inner part being connected to one another at a fixed distance; andat least one cone sliding part movably arranged between the inner part and the outer part. 2. The collimator according to claim 1, wherein all conical surfaces are arranged concentrically about a common axis of rotation. 3. The collimator according to claim 1, wherein all conical surfaces have the same aperture angle α. 4. The collimator according to claim 1, wherein pairs of adjacent conical surfaces have the same aperture angle. 5. The collimator according to claim 1, wherein the at least one cone sliding part travels in a direction of an axis of rotation. 6. A method for adjusting a focal length of a collimator, the method comprisingproviding a collimator for an X-ray system, the collimator having an outer part having a conical inner surface, an inner part having a conical outer surface, the outer part and the inner part being connected to one another at a fixed distance, and at least one cone sliding part movably arranged between the inner part and the outer part; andmoving the at least one cone sliding part along an axis of rotation until a desired focal length is achieved,wherein, when multiple cone sliding parts are used, they are moved independently of one another. 7. An x-ray inspection system comprising:an X-ray source;an X-ray detector;an X-ray detector and an analysis device for analyzing detected radiation; anda collimator with an adjustable focal length, the collimator comprising:an outer part having a conical inner surface;an inner part having a conical outer surface, the outer part and the inner part being connected to one another at a fixed distance; andat least one cone sliding part movably arranged between the inner part and the outer part. |
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051456362 | description | EXAMPLES Example 1 Preparation of Aluminum Perrhenate Soluble Irradiation Target Procedure No. 1 0.1 ml concentrated nitric acid was carefully pipetted into a beaker containing 50 mg rhenium metal. After two to three minutes, 0.1 ml 1:1 v/v concentrated nitric acid:water solution was added to the reaction mixture, and the mixture was allowed to react at room temperature for 20 minutes. 0.5 ml water was added to the mixture, followed by the addition of 0.2 ml of a 0.447M aluminum chloride solution. The mixture was heated to dryness at 95.degree. C. for 12 hours. 1 ml water was added to the beaker to dissolve the residue solid, then the solution was filtered. The filtrate was heated to dryness at 95.degree. C. for 12 hours to yield a white solid, aluminum perrhenate. The elemental analysis of Al and Re is shown in Table I. Procedure No. 2 To 50 mg rhenium metal was carefully added 0.5 ml 30% H.sub.2 O.sub.2 solution. After 10 minutes, 0.5 ml water was added to the mixture and the mixture was heated for 30 minutes at 95.degree. C. 0.2 ml of 0.447M aluminum chloride solution was added to the mixture, and the mixture was heated to dryness at 95.degree. C. for 12 hours. 1 ml water was added to dissolve the solid, then the solution was filtered. The filtrate was heated to dryness at 95.degree. C. for 12 hours to obtain a white solid, aluminum perrhenate. Procedure No. 3 To 50 mg rhenium metal was carefully pipetted 0.1 ml concentrated nitric acid. After two to three minutes, 0.1 ml 1:1 v/v concentrated nitric acid:water solution was added to the reaction mixture, and the mixture was allowed to react at room temperature for 20 minutes. 0.1 ml ammonium hydroxide (conc.) was added to the reaction mixture, followed by the addition of 0.5 ml water. The mixture was heated at 95.degree. C. until all solids had dissolved, and then 0.2 ml of 0.447M aluminum chloride solution was added to the mixture. The solution was heated to dryness at 95.degree. C. for 12 hours. 1 ml water was then added to dissolve the solid, and the solution was filtered. Finally, the filtrate was heated to dryness at 95.degree. C. for 12 hours to yield a white solid, aluminum perrhenate. Procedure No. 4 The procedure of Procedure No. 1 was followed, except 50 mg enriched Re-185 rhenium metal (95% Re-185, Isotec, Inc., Dayton, Ohio) was used in place of the natural occurrence rhenium metal of Procedure 1. Procedure No. 5 The procedure of Procedure No. 2 was followed, except 75 mg enriched Re-185 rhenium metal (95% Re-185, Isotec, Inc., Dayton Ohio) and 0.3 ml 0.477M aluminum chloride solution was used in place of the natural occurrence rhenium metal and aluminum chloride solution of Procedure 2. Procedure No. 6 The procedure of Procedure No. 3 was followed, except 50 mg enriched Re-185 rhenium metal (95% Re-185, Isotec, Inc., Dayton, Ohio) was used in place of the natural occurrence rhenium metal of Procedure 3. The products of Procedure Nos. 1-6 were dissolved in 2 ml of water, diluted 1:200 in water, analyzed for rhenium and aluminum content and found to contain the amounts of rhenium and aluminum shown in the following Table 1 as .mu.g per ml of sample. TABLE I ______________________________________ Elemental Content (.mu.g/ml) Procedure Number #1 #2 #3 #4 #5 #6 ______________________________________ Al 5.0 5.2 5.4 5.1 7.6 4.5 Re-total 125.8 116.7 131.4 104.6 157.4 103.6 Re-185 46.1 42.6 48.1 97.7 148.4 97.5 Re-187 79.7 74.1 83.2 7.2 9.0 6.1 Re-Al ratio 25.2 22.4 24.3 20.5 20.7 23.0 ______________________________________ EXAMPLE II Irradiation of Soluble Aluminum Perrhenate Targets 0.24 mg samples of aluminum perrhenate produced by Procedures 4, 5 and 6 of Example 1 were placed in quartz irradiation vials, which were sealed by flame under vacuum. The vials were irradiated in a flux of 3.times.10.sup.14 neutrons/cm.sup.2 /sec at the University of Missouri Research Reactor (Columbia, Mo.) for a period of 12-14 days. The vials were then opened and the irradiated aluminum perrhenate was dissolved in 0.5 ml of water. The average activity recovery was found to be greater than 90%, and the activity was identified as .sup.186 Re (by germanium-lithium analyzer) in the form of .sup.186 ReO.sub.4.sup.- (by reverse phase HPLC). EXAMPLE III Preparation of Sodium Tungstate Target 0.5 ml of a 5N NaOH solution was added to 0.2 g of .sup.186 WO.sub.3 (96% .sup.186 W, Oak Ridge National Laboratory, Oak Ridge, Tenn.) in a 5 ml beaker, and the mixture was heated to dissolve the .sup.186 WO.sub.3 material. After dissolution was complete, the mixture cooled to room temperature and adjusted to pH 9 by the addition of 5N HCL and 1N HCl. The solution was heated to dryness to obtain 0.292 g of a white solid, comprising sodium tungstate. EXAMPLE IV Irradiation of Soluble Sodium Tungstate Target A 20 mg sample of sodium tungstate (W-186) prepared by the procedure of Example III was placed in a quartz irradiated vial, which was sealed by flame under vacuum. The vial was irradiated in a flux of 3.times.10.sup.14 neutrons/cm.sup.2 /sec at the University of Missouri Research Reactor (Columbia, Mo.), for a period of seven days. .sup.187 W and .sup.24 Na were allowed to decay from the irradiated material. The vial was then opened and the irradiated sodium tungstate was dissolved in 1 ml of water. The solution was then filtered through a 0.2 .mu.m filter. The activity recovery was determined to be greater than 95%. EXAMPLE V Preparation of Zirconium Tungstate Generator 40 mg of irradiated sodium tungstate prepared as described in Example IV was dissolved in 2 ml of water and filtered through a 0.2 .mu.m filter. To the filtered solution 200 mg of natural occurrence Na.sub.2 WO.sub.4 in 4 ml H.sub.2 O was added to aid in precipitation of insoluble tungstate. The tungstate solution was then added drop by drop to an acidic zirconyl solution containing 312 mg ZrO(NO.sub.3).sub.2 in 6 ml of HCl solution, pH 1. The pH of the resulting zirconium tungstate mixture was 6.0. The mixture was then filtered and washed with 50 ml of water. The gel-precipitation yield (percentage of .sup.188 W initially present which is complexed in the gel) was greater than 95%. The filtered gel was air dried overnight, packed into a conventional generator column, and eluted with physiological saline. The zirconyl tungstate generator was determined to have an elution yield of 66%. While the invention has been described in connection with various presently particularly preferred embodiments, various modifications will be apparent to those skilled in the art. Any such modifications are intended to be within the scope of the appended claims, except insofar as precluded by the prior art. |
abstract | A lithographic apparatus including a filter device is disclosed. The filter device has a plurality of foils attached to a holder which is able to rotate around a rotation axis. The foils are arranged substantially parallel to the rotation axis. The foils comprise a uni-directional carbon-fiber composite material selected from the group consisting of carbon-carbon composite (C-C composite) and carbon-silicon carbide composite (C—SiC composite). During operation, the filter device rotates and filters out debris from a radiation source, such as a Sn plasma source. Such a filter device per se may be provided. |
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056278667 | abstract | A fuel assembly for a nuclear reactor vessel includes a plurality of fuel rods, at least one coolant rod, a lower tie plate supporting the fuel rods and coolant rod, and a channel surrounding the fuel rods, coolant rod and tie plate. The lower tie plate is supported by the channel such that the channel carries a load of the fuel assembly. In one aspect of the invention, an upper tie plate includes two spring loaded latch pins engageable with corresponding apertures in the channel. A transition member supports the lower tie plate and is rigidly secured to the channel. Thus, when lifting the fuel assembly from the reactor, the channel bears the load of the fuel assembly. End gussets or clips are welded to the channel and inserted into the transition member, which serve as a secondary support for the transition member in the event that its primary connection to the channel fails. A channel guide member is secured to the, upper tie plate and includes two ears that are received in ear apertures in the channel. The channel guide and ears thus provide a redundant attachment between the tie plate and the channel in the event that the spring loaded latch pins fail. If it is desired to remove the fuel bundle from the channel, the upper tie plate is released from the assembly by detaching the guide member and releasing the spring loaded latch pins. The bundle can then be removed from the channel by attaching a grapple head to the coolant rod ends, which are specially shaped to facilitate an attachment tool. |
abstract | A separable multi-component cask for spent nuclear fuel transport and storage includes a vertically elongated outer cylinder having a neutron radiation shielding composition and a vertically elongated inner cylinder having a gamma radiation blocking composition. The inner cylinder includes a cavity configured to hold a spent nuclear fuel canister. The inner cylinder is detachably mounted and nested inside a cavity of the outer cylinder and is separable therefrom during spent fuel cask loading operations in a staged manner. An air ventilation annulus formed between the first and second cylinders forms a heat removal passage to remove heat emitted by the radioactive canister when placed inside the second cylinder. A pair of removably coupled mating top flanges on the inner and outer cylinders supports and suspends the inner cylinder in a cantilevered manner, thereby allowing the directly heated inner cylinder to thermally expand to a greater degree than the outer cylinder. |
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description | This application claims priority to U.S. Provisional Application Ser. No. 61/368,762 filed on Jul. 29, 2010, the specification of which is herein incorporated by reference in its entirety. Radioactive molybdenum (99Mo) is used to produce technetium (99mTc), which is an ingredient for a wide range of radiopharmaceuticals used in the health care industry. A continuous supply of the technetium isotope is needed for approximately 80% of all nuclear medicine procedures worldwide, including nearly 50,000 diagnostic procedures per day in the U.S. alone. The demand may continue to grow as the world's population ages. Conventional production of molybdenum occurs in high power reactors, e.g., nuclear reactors which produce in excess of ten megawatts of thermal energy. Current regulations may allow for a limited or restricted use of the high power reactors utilizing highly-enriched uranium (HEU) for generation of isotopes such as molybdenum. However, the heightened scrutiny due to concerns over proliferation of HEU may ultimately limit or prohibit the use of this material for the production of molybdenum and other radioactive isotopes. Countries such as the United States may find themselves in a compromised position concerning the future production of molybdenum. The sole source of molybdenum in North America is presently located in Canada, and there is growing concern that the continued availability of molybdenum from this source may end in the near future. Although there are other sources in Europe, the half life of molybdenum is sufficiently short, (2.75 days), making transportation around the world an untenable solution. Accordingly, for many countries, local production of molybdenum may be the only viable long term option. FIG. 1 illustrates an example isotope target 10. The isotope target 10 may be used to produce a radioactive isotope such as molybdenum (99Mo). The isotope target 10 is illustrated as being cylindrical in shape, with an outer diameter wall 3 and an inner diameter wall 9. However, targets having other shapes are contemplated herein, including hexagonal cross-sections and other geometries. The outer diameter wall 3 may be associated with a first diameter, and the inner diameter wall 9 may be associated with a second diameter. The first diameter is greater than the second diameter. The isotope target 10 may comprise a central region 15 which extends from a first end 12 of the isotope target 10 to a second end 14 of the isotope target 10. The central region 15 may comprise a hollow portion, a channel, a cavity, a through-hole, a tube, or the like. FIG. 2 illustrates a cross-section 20 of the example isotope target 10 of FIG. 1. The isotope target 10 may comprise a first tube 2 and a second tube 4. The second tube 4 may be nested with the first tube 2 to form a target chamber 1. The first tube 2 may comprise the outer diameter wall 3 of the isotope target 10 and an inside wall 5. The second tube 4 may comprise the inner diameter wall 9 of the isotope target 10 and an outside wall 7. The target chamber 1 may be located between the inside wall 5 of the first tube 2 and the outside wall 7 of the second tube 4. In one example, the target chamber 1 may extend substantially the entire length of the isotope target 10. The target chamber 1 may be sealed at the first end 12 (FIG. 1) and the second end 14 (FIG. 1) of the isotope target 10. Additionally, a target material including an isotope source or other radioactive and/or fissile material, such as uranium, e.g., 235U, may be located in the target chamber 1. The target material may be located between the inner diameter wall 9 and the outer diameter wall 3. In one example, the target material may be interspersed with one or more voided regions. The central region 15 may be located within the inner diameter wall 9. In one example, the central region 15 may be configured to house a neutron thermalization volume. For example, the neutron thermalization volume may comprise water, heavy water, graphite, zirconium, plastic, wax, paraffin, hydrogenous materials, other types of neutron moderators, or any combination thereof. The central region 15 may form a water channel configured to allow water to flow through the isotope target 10. For example, water may enter the central region 15 through the first end 12 (FIG. 1) and exit the central region 15 at the second end 14 (FIG. 1). In another example, the central region 15 may comprise a completely enclosed chamber configured to house the neutron thermalization volume. The central region 15 may be configured to cause neutrons that are generated in the target material located in the target chamber 1 to be thermalized by the neutron thermalization volume before re-entering the target material. The neutron thermalization volume, e.g., water or primary coolant, may also be used to remove heat from and/or cool the isotope target 10 during fission events and/or during an isotope production process. The target material may be located within the target chamber 1 in a variety of different geometries. FIG. 3 illustrates an example isotope production target 30 shown, by way of example, with a cross-sectional view of an isotope target, similar to isotope target 10 illustrated in FIG. 1. The isotope production target 30 may comprise an outer cladding 32 and an inner cladding 34. Target material 31 may be located between the outer cladding 32 and the inner cladding 34. The target material 31 may comprise fissile material 36 and one or more voided regions 38. The one or more voided regions 38 may be configured to capture fission product gases produced from the fissile material 36. Capturing the fission product gases in the one or more voided regions 38 may reduce the amount of fission product gases that become interstitial and which may otherwise cause structural degradation of the surrounding cladding in a target which does not include any voiding between the fissile material and the cladding. In one example, the fissile material 36 may comprise fissile source pellets, and the one or more voided regions 38 may comprise spacing or gaps between the pellets. The fissile material 36 may comprise a plurality of individual source objects, may be stored in a powder form, or take other physical forms such as balls, fragments, particles, sheets, rods, foils, other geometries, or any combination thereof. The one or more voided regions 38 may be sealed to prevent the fission gases from exiting the isotope production target 30. For example, the target material 31 may be contained in a sealed chamber located between the outer cladding 32, the inner cladding 34, and the ends of the isotope production target 30, such as the first end 12 and the second end 14 of the isotope target 10 illustrated in FIG. 1. The voided regions 38 may comprise one or more gases, a vacuum, or a partial vacuum, e.g., prior to capturing any fission product gases. A central region 35 of the isotope production target 30 may comprise a neutron thermalization volume or neutron moderator. The neutron thermalization volume may comprise water. In one example, the isotope production target 30 may be configured to be installed in a reactor core, and the neutron thermalization volume may comprise a primary coolant associated with the reactor core. The reactor core may be associated with a low power reactor with less than ten megawatts of thermal output. For example, a low power reactor such as a Training, Research, Isotopes, General Atomics, or TRIGA®, reactor may be used to produce certain isotopes, such as molybdenum. The isotope production may be accomplished through a series of operations or generalized steps. In a first operation, a suitable isotope production target may be manufactured. Manufacture of an isotope production target, such as the isotope production target 30, may comprise placing a target material, such as uranium, in a particular geometry within the isotope production target. In a second operation, the isotope production target may be irradiated by a neutron source. For example, the isotope production target may be placed in a nuclear reactor. During irradiation, fission reactions in the target material may produce one or more isotopes, such as molybdenum. Fission gases or by-products are also typically generated during the fission reactions. In one example, the fission gases and/or by-products may be captured or stored within voided regions interspersed with the target material. In a third operation, the isotope production target may be transported to a hot cell facility for remote handling. Inside the hot cell, the irradiated target material may be removed from the cladding. In one example, an end of the isotope production target, such as the first end 12 illustrated in FIG. 1, may be cut or otherwise removed to extract the target material. For example, the fissile material 36 illustrated in FIG. 3 may be loosely placed within the chamber 1 (FIG. 2) for easy removal, e.g., by inverting the isotope target 10 with the first end 12 removed. A series of chemical separations of the target material may be performed to produce or extract the desired end product, such as pure molybdenum. In a fourth operation, the end product may be transported to a destination such as a distribution facility, a hospital, a clinic, a laboratory, a test facility, a research facility, a place of business, a governmental facility, or the like. In one example, technetium (99mTc) that is obtained from the end product, e.g., molybdenum, may be used for medical procedures at the destination. FIG. 4 illustrates an example target 40 and a central region 45. The target 40 may comprise a first end 42 and a second end 44. In one example, one or both of the first end 42 and the second end 44 may be removed, e.g., after the target 40 has been irradiated. The target may be configured to have an outside diameter of approximately 1.43 inches (3.63 centimeters) and a height of approximately 22 inches (171.63 centimeters). The target 40 may be configured to approximate the overall dimensions of a fuel element for a TRIGA® reactor, or other type of reactor. FIG. 5 illustrates a cross-section 50 of the example target 40 of FIG. 4 taken at or near the first end 42. The target 40 may comprise an outer cladding 52, an inner cladding 54, and a target chamber 51 formed there between. Fissile material may be located within the chamber 51. In one example, the fissile material may comprise two layers of fissile material, including a first layer 53 and a second layer 56. A voided region 58 may be located between the first layer 53 and the second layer 56. The voided region 58 may comprise an annulus, or be annular in shape. The voided region 58 may be filled with a gas, or gases, or may be configured as a vacuum, or a partial vacuum. The voided region 58 may be configured to operate as storage or a volume for collecting fission product gases and/or by-products. Initially providing the voided region 58 with the vacuum, or partial vacuum, may allow for the collection of a greater amount of the fission product gases generated during irradiation of the fissile material, in order to further reduce an overall pressure within the voided region 58 during an isotope production operation. The outer cladding 52 and the inner cladding 54 may comprise two nested and/or sealed tubes. The top and bottom of the nested tubes may be sealed such that fission gases produced during irradiation may be trapped in the voided region 58. The outer cladding 52 and/or the inner cladding 54 may be made of stainless steel, aluminum, and/or other materials, and may be manufactured with a thickness that is nominally 0.020 inches (0.06 centimeters). The precise thickness of the fissile material and cladding may vary depending on various design considerations, such as available neutron flux, production yield requirements, material characteristics, reactor core geometry, or any combination thereof. The inner cladding 54 may be configured as a channel or a container for a neutron moderator 55. The neutron moderator 55 may be located within the inner cladding 54 and may be configured to cause neutrons that are generated in the fissile material, e.g., the first layer 53 and/or the second layer 56, to be thermalized by the neutron moderator 55 before re-entering the fissile material. The thermalized neutrons may be used to produce additional fission events in the first layer 53 and/or the second layer 56. In one example, the neutron moderator 55 may comprise graphite, zirconium, plastic, wax, parafin, hydrogenous materials, other types of neutron moderators, or any combination thereof. In another example, the neutron moderator 55 may comprise water, such as light water or heavy water, which is allowed to flow through the channel formed within the inner cladding 54 during an isotope production operation. The neutron moderator 55 may comprise primary coolant from a reactor. The outer cladding 52 and/or the inner cladding 54 may keep the first layer 53 and the second layer 56 from contacting any water or primary coolant. In one example, a hole may be opened, e.g., punched, in the side of the target 40 and the fission gases and/or by-products may be extracted from the voided region 58 to be collected and/or stored. One or both ends of the target 40, e.g., the first end 42 and/or the second end 44 (FIG. 4), may be removed or cut. The end product, e.g., molybdenum, may be extracted from the first layer 53 and/or the second layer 56. For example, the end product may be chemically separated from the irradiated material. The rate of fission reactions in an isotope production target may be described by the equation:R=φσN, where R=reaction rate density of fission [fissions cm−3 s−1] φ=neutron flux from reactor [neutrons cm−2 s−1] σ=microscopic cross-section for fission [cm−2] N=atomic density of target atoms [atoms cm−3] Neutrons and fission fragments may be produced directly from fission events. About 6.5% of the time, the isotope molybdenum may be created as a fission fragment of an 235U target irradiated with thermal neutrons. The above equation may describe a fission rate density in a fissile material. In order to maximize the fission rate density, the values of flux and atomic density in the equation may be changed; the microscopic cross-section is a fixed parameter. Many types of research reactors and low power reactors may be associated with a nominal power of one megawatt thermal (MWt) and may have neutron fluxes on the order of 1E13 neutrons cm−2 s−1. With the cross-section fixed, the atomic density N and/or the neutron flux φ may be increased by configuring the geometry and/or the materials of the target as described herein, with reference to the various examples. Although the flux of neutrons emanating from the reactor core may be associated with a fixed value in some examples, the geometry of the target may be used to increase the flux of neutrons within the target itself. The neutron flux φ may comprise the flux of neutrons from both the reactor core and from the target, e.g., neutrons generated within the target. Neutrons born from fission events in the target may have the opportunity to thermalize within the neutron moderator, such as water, located within the target, and the thermal neutrons may continue on to create more fission reactions in the target. In one example, substantially all of the fissile material, e.g., uranium, may be located on one layer (e.g., on the inside surface of the outer cladding 52). However, having two layers of fissile material, e.g., the first layer 53 and the second layer 56, may provide for improved removal of heat from the fission reactions. For example, the amount of heat removed from the target 40 may correspond to the amount of surface area of the cladding that is in direct contact with the neutron moderator 55, e.g., water. In one example, both the outer wall of the outer cladding 52 and the inner wall of the inner cladding 54 may be exposed to water, which may cool the surface(s) of the target 40. To increase the atomic density (N) of the target material, e.g., uranium, the target 40 may be configured with uranium metal which has a density of approximately 18 g cm−3 or nearly four times the density of uranium oxide (UO2). A higher density results in a higher atomic density N. In one example, the target material may comprise low-enriched uranium (LEU), enriched to approximately 19.75%. Fissile material which is enriched at or above 20% may be termed or defined as HEU, and fissile material which is enriched below 20% may be termed or defined as LEU. The thickness of the first layer 53 and/or the second layer 56 may be allowed to vary depending upon the desired mass of target material. The greater the mass, the greater the value of N and therefore the greater the production rate of molybdenum. In one example, the first layer 53 and/or the second layer 56 may be “sputtered” onto, or otherwise adhered to, the outer wall 52 and the inner wall 54, respectively. The first layer 53 may have a different mass compared to the second layer 56 due to a difference between their radial location, e.g., cylindrical geometry, and/or thickness. In one example, the first layer 53 and the second layer 56 may be loosely fit within the target chamber 51, e.g., not adhered to either the inner wall 54 or the outer wall 56, respectively. Accordingly, the first layer 53 and the second layer 56 may be physically removed from the target 40 without performing any chemical or thermal treatment. In another example, a chemical may be inserted or injected into the voided region 58 after the fissile material has been irradiated, in order to dissolve the first layer 53 and the second layer 56, for removal from the cladding. FIG. 6 illustrates a table 60 showing an example isotope production rate. Table 60 illustrates the production of molybdenum as a function of target mass, e.g., uranium. A relative change in isotope production may be determined as a function of the mass of the fissile material within the target. The target may comprise a mass of fissile material selected somewhere between 200 and 400 grams; although other masses of fissile material may be used. In one example 62, a target comprising a mass of 200 grams of uranium may produce approximately 300 Curies (Ci) of molybdenum (99Mo), and in another example 64, a target comprising a mass of 400 grams of uranium may produce approximately 450 Curies of molybdenum. As the mass of fissile material is increased, the amount of the end product, such as molybdenum, may also increase. This increase in amount of the end product may not be linear though because the neutron flux may diminish as it penetrates the fissile material. This phenomenon may be termed or known as self-shielding. As a result, the example isotope production curve illustrated in FIG. 6 may approach or reach a maximum value instead of continuing to increase linearly with mass of the fissile material. The mass of the fissile material may be varied by increasing or decreasing a diameter, a thickness, a length, a width, a height, a composition, or any combination thereof, associated with the fissile material. FIG. 7 illustrates an example target assembly 70 comprising an isotope production target 76. In one example, the isotope production target 76 may be approximately sized as a fuel element of a reactor core. A mounting structure 77 may be coupled to the isotope production target 76 and may be configured for insertion of the isotope production target 76 into a reactor core. A first portion 71 of the mounting structure 77 may be coupled to the isotope production target 76 at a first end 72, and a second portion 73 of the mounting structure 77 may be coupled to the isotope production target 76 at a second end 74. The mounting structure 77 may comprise one or more holes 75. The one or more holes 75 (hereafter “holes”) may be configured to direct water or primary coolant into or through the target assembly 70. The holes 75 may be located about a circumference of one or both of the first portion 71 and the second portion 73. The holes 75 may be configured to provide a path for water or primary coolant to enter into, or exit out of, the target assembly 70. FIG. 8 illustrates a cross-section 80 of the example target assembly 70 of FIG. 7. The isotope production target 76 may comprise an outer wall 82 and an inner wall 84. The isotope production target 76 may be configured to contain fissile material in an isotope production chamber located between the outer wall 82 and the inner wall 84. Additionally, the isotope production target 76 may comprise a central region 85 located within the inner wall 84. In one example, the length 86 of the fissile material included in the isotope production target 76 may be approximately 20 inches. The mounting structure 77 may be configured to direct primary coolant associated with a reactor core to pass through the central region 85. In addition to, or in place of, holes 75 the mounting structure 77 may comprise a first opening 87 located in or near the first portion 71, and a second opening 88 located in or near the second portion 73. One or both of the first opening 87 and the second opening 88 may be configured to allow water or primary coolant to pass into, or out of, the central region 85. The central region 85 may be configured to thermalize neutrons generated by the fissile material when the isotope production target 76 is inserted into the reactor core. The mounting structure 77 may comprise a connecting device 89. The connecting device 89 may be configured to couple the mounting structure 77 to the isotope production target 76. Each of the first portion 71 and the second portion 73 of the mounting structure may be coupled to the isotope production target 76 by a connecting device, such as the connecting device 89. FIG. 9 illustrates an exploded view of an isotope production assembly 90, including an isotope target structure 95, a first mounting structure 92, and a second mounting structure 94. The first mounting structure 92 may comprise one or more openings 97 configured to allow coolant to flow into and/or out of the isotope target structure 95. The second mounting structure 94 also may comprise one or more openings. The first mounting structure 92 and the second mounting structure 94 each may comprise a target insert 96. The target insert 96 may comprise a connecting device 93. In one example, the connecting device 93 may be configured to connect the target insert 96 to the first mounting structure 92 and/or to the isotope target structure 95. FIG. 10 illustrates a further cross-section 100 of the example target assembly 70 of FIG. 7. The mounting structure 77 may be connected to the isotope production target 76 by connecting device 89. The connecting device 89 may comprise one or more support arms, flutes, webbing, or the like. The support arms may radiate outward from the mounting structure 77 to connect to the isotope production target 76. One or more openings, such as opening 105 may be formed between or through the connecting device 89. In one example, the mounting structure 77 may be configured to direct water or primary coolant through the opening 105 into the isotope production target 76. Water or primary coolant passing through the isotope production target 76 may be allowed to exit the opening 105. An isotope production chamber 101 may be located between the outer wall 82 and the inner wall 84, and may be configured to house fissile material. Neutrons generated in the fissile material may be thermalized, or moderated, by the water or primary coolant entering and/or exiting the isotope production target 76 before re-entering the isotope production chamber 101. The neutrons that re-enter the fissile material may cause additional fission events that may generate further neutrons that may then be thermalized by the primary coolant in the central region of the isotope production target 76. FIG. 11 illustrates a target 110 comprising a vacuum chamber 115. The target 110 may comprise a single, thin-walled stainless steel tube 112 coated on the inside with uranium oxide (UO2) 114. The uranium oxide 114 may comprise HEU with an enrichment of approximately 93% and a density of approximately 4.8 g cm−3. The illustrated target 110 may be similar to targets used in a so called “Centichem” process utilizing a high power reactor. Any neutrons produced from fission in the uranium oxide 114 may not have the opportunity to thermalize in the vacuum chamber 115 while the neutrons remain in the target 110. Accordingly, the probability that these high energy neutrons will produce another fission event of the uranium oxide 114 may be extremely low. The vast majority of neutrons produced from the fission events may simply leak out of the target. Instead, the target 110 may have to rely on neutrons entering the stainless steel tube 112 from the outside of the target 110, e.g., neutrons born from a remote neutron source. Accordingly, the neutron flux associated with the target 110 may be less than the neutron flux associated with a target, such as the isotope target 10 of FIG. 1, which includes a central region configured to house a neutron thermalization volume. Coating the uranium oxide on the inside of the stainless steel tube 112 may require processing the material in a chemical bath. The chemical bath may be used to dissolve both the uranium oxide 114 and the stainless steel tube 112, which may complicate the separation and processing of the desired isotopes. There may be no space or voiding between the uranium oxide 114 and the stainless steel tube 112. FIG. 12 illustrates an example lattice configuration 120. The lattice configuration 120 may comprise a grid plate for a reactor core assembly comprising a plurality of fuel rods and one or more targets. In one example, the lattice configuration 120 may comprise a number of concentric rings of fuel rods. In the illustrated example, an outer ring 122, or “G-ring”, may comprise thirty-six positions; an “F-ring” 124 may comprise thirty positions; an “E-ring” 126 may comprise twenty-four positions; a “D-ring” 128 may comprise eighteen positions, etc. A central position, or “A-ring” may comprise a single position. One or more targets may be located at any of the positions of the lattice configuration 120. In one example, a target may be located in a position associated with the outer ring 122, in order to facilitate access, e.g., installation and/or retrieval, of the target. The position of the target may also be used to control the neutron flux received by the target and/or the heat generated by the target. In order to increase the power density and/or the neutron flux, the target may be moved closer to the center of the lattice configuration 120. In one example, approximately two fuel rods may be removed from the lattice configuration 120 for every three targets added. One or more targets may be added to the lattice configuration 120 while maintaining the overall design characteristics, e.g., certification and operating criteria, of the reactor. The target may be placed in the reactor core for a number of hours or days, e.g., six days, during the isotope production process. FIG. 13 illustrates a table 130 showing a comparison between various targets using an example moderator of water. The illustrated table 130 compares the production rate of molybdenum as a function of position, or element number, of a target located in the outer ring 122 of the lattice configuration 120 of FIG. 12. The molybdenum production rate for a first target 132 comprising beryllium cladding and a central region comprising a neutron thermalization volume or neutron moderator of water, such as the central region 15 of FIG. 1, is illustrated as varying between approximately 280 Curies (Ci) and 405 Curies. The molybdenum production rate for a second target 134 comprising stainless steel cladding and a central region comprising a neutron thermalization volume or neutron moderator of water is illustrated as varying between approximately 230 Curies and 300 Curies. Both the first target 132 and the second target 134 may comprise 200 grams of fissile material, for purposes of the present illustration and comparison. The molybdenum production rate for a third target 136 comprising 200 grams of fissile material and a vacuum chamber, such as the vacuum chamber 115 of FIG. 11, is illustrated as varying between approximately 160 Curies and 230 Curies. The molybdenum production rate for a fourth target 138 comprising 100 grams of fissile material and a vacuum chamber is illustrated as varying between approximately 105 Curies and 150 Curies. The first target 132 and the second target 134 may generally be understood to provide for a higher isotope production rate than either the third target 136 or the fourth target 138. By including a central region comprising a neutron thermalization volume or neutron moderator, the first target 132 and the second target 134 may be able to more efficiently utilize the available neutrons, e.g., neutron flux, by effectively increasing the number of thermal neutrons in the target, and thereby increasing the fission rate of the fissile material. Including a central neutron thermalization volume in the target may provide for an improved and more cost-effective method for the production of molybdenum and other isotopes, such as plutonium. For the same neutron flux, some example targets comprising the central neutron thermalization volume may produce approximately three times the amount of molybdenum in a low power reactor compared to a target comprising a vacuum chamber. Other types of cladding may be utilized in various example targets, including zirconium, zirc-alloy, aluminum, ceramics, other materials, or any combination thereof. FIG. 14 illustrates an example target 140 and a multi-layered source structure. The target 140 is illustrated as comprising a first cladding assembly 142 and a second cladding assembly 144. The second cladding assembly 144 may be nested within the first cladding assembly 142. The first cladding assembly 142 may comprise a first source 141 of fissile material, and the second cladding assembly 144 may comprise a second source 143 of fissile material. The geometry of the multi-layer source structure may be used to increase the mass of the fissile material as compared to a target comprising a single layer of fissile material, for example. A central region 145 of the target 140 may be located within an inner wall 148 of the second cladding assembly 144. The central region 145 may comprise a neutron thermalization volume or neutron moderator, such as water or primary coolant. Additionally, an intermediate region 147 may comprise a neutron thermalization volume or neutron moderator, such as water or primary coolant. The intermediate region 147 may be located outside of an outer wall 149 of the second cladding assembly 144, for example between the first cladding assembly 142 and the second cladding assembly 144. In one example, water or other types of primary coolant may be allowed to flow through the central region 145 and/or the intermediate region 147, such that there may be two or more channels of water flowing through the target 140. Furthermore, including the intermediate region 147 between the first source 141 of fissile material and the second source 143 of fissile material may reduce the effect of self-shielding due to the increased mass of fissile material in the target 140. The rate of neutron thermalization and/or the rate of isotope production may be controlled by configuring the target 140 to vary the amount of fissile material and/or to adjust the volume and/or the rate of flow of the water through the target 140. Example Modes of Operation FIG. 15 illustrates an example process 150 of isotope production. At operation 151, fissile material located between an outer wall and an inner wall of an isotope production target may be stored. A fissile material housing may comprise the outer wall and the inner wall. The fissile material may be stored in a target chamber located between the outer wall and the inner wall. At operation 152, primary coolant, such as water, may be directed through the isotope production target. The primary coolant may be directed through a central region or channel of the isotope production target. In one example, the primary coolant may be directed through the isotope production target by a mounting device located at one or both ends of the isotope production target. At operation 153, neutrons generated during fission events of the fissile material may be thermalized in a central region of the isotope production target. The neutrons may be generated in response to irradiating the fissile material with a neutron source. The central region may comprise a neutron thermalization volume or neutron moderator, such as light water, heavy water, graphite, zirconium, plastic, wax, paraffin, hydrogenous materials, other types of neutron moderators, or any combination thereof. In one example, the central region may comprise primary coolant associated with a reactor core. The central region may comprise water that thermalizes the neutrons generated during the fission events, and the thermalized neutrons may cause additional fission events of the irradiated material. In one example, the number of fission events may be approximately doubled due to the thermalized neutrons versus a target which is only irradiated with neutrons born from the neutron source located outside of the target. At operation 154, fission by-products generated during fission events of the irradiated material may be captured in one or more voided regions interspersed with the fissile material. The one or more voided regions may be located between the outer wall and the inner wall of the isotope production target. In one example, the one or more voided regions may be interspersed between a plurality of objects, such as balls or pellets, comprising the fissile material. In another example, the fission by-products may be stored in an annulus located between two layers or sheets of the fissile material. At operation 155, the irradiated material may be removed from the isotope production target. In one example, one or more ends of the isotope production target may be removed, e.g., cut, prior to removing the irradiated material. Fissile material stored as a plurality of objects may be loosely contained in the isotope production target and the irradiated material may be physically removed from the isotope production target without any chemical or thermal treatment. In one example, the fissile material may be stored as one or more sheets, foils, tubes, or the like, within the target chamber. The fissile material may adhere, e.g., be sputtered, on the inner wall and/or the outer wall of the target chamber. At operation 156, the irradiated material may be chemically treated to separate the isotopes, e.g., molybdenum isotopes. The irradiated material may be treated in a chemical bath or an acid bath, for example. In one example, the irradiated material may be chemically treated after the irradiated material has been removed from the isotope production target. In another example, the irradiated material may be chemically treated while it remains in the isotope production target. For the sake of convenience, the operations may be described as various interconnected functional blocks or diagrams. This is not necessary, however, and there may be cases where these functional blocks or diagrams are equivalently aggregated into a single operation with unclear boundaries, and/or where one or more of the operations may be omitted from the process. Whereas certain examples have described using the target in a low power reactor, such as a TRIGA® reactor, one skilled in the art would appreciate that the target may also be used in plate fuel-type research reactors or in high power reactors, for example with power capacity greater than ten megawatts thermal. Whereas various examples may be described with the target comprising LEU, other examples may include HEU, uranium oxide UO2, plutonium, 233U, or any combination thereof. Having described and illustrated the principles of various examples, it should be apparent that the examples may be modified in arrangement and detail without departing from such principles. I claim all modifications and variations coming within the spirit and scope of the following claims. |
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claims | 1. A spacer grid for a nuclear fuel assembly, the spacer grid comprising:a plurality of grid strips assembled in a lattice pattern; anda plurality of grid cells formed by the plurality of grid strips,wherein a section of each of the grid strips, corresponding to each grid cell of the plurality of grid cells, includesa dimple supporting a fuel rod and having slits formed in a lateral direction perpendicular to the fuel rod and arc-shape portions protruding inwards or outwards between the slits, andat least one set of slots formed entirely in a planar surface of the section separately from the slits surrounding the dimple and having at least one of a wave shape, a saw tooth shape and a slanted line shape extending in the lateral direction along the slits of the dimple, said at least one set of slots including a first slot disposed in an upper position of the section above and apart from the dimple and a second slot disposed in a lower position of the section below and apart from the dimple,wherein said at least one set of slots is provided for reducing flow-induced high-frequency vibration of the spacer grid. 2. The spacer grid as set forth in claim 1, wherein the first slot and the second slot have the same shape each other. 3. The spacer grid as set forth in claim 1, wherein the first slot and the second slot have different shapes each other. 4. The spacer grid as set forth in claim 1, wherein the first slot and the second slot are vertically symmetric with respect to an imaginary center line of the section. 5. The spacer grid as set forth in claim 1, wherein the at least one set of slots further includes a third slot disposed in the upper position of the section apart from the dimple and a fourth slot disposed in the lower position of the section apart from the dimple. 6. The spacer grid as set forth in claim 5, wherein the third slot and the fourth slot have the same shape each other. 7. The spacer grid as set forth in claim 5, wherein the third slot and the fourth slot have different shapes each other. 8. The spacer grid as set forth in claim 5, wherein the third slot and the fourth slot are vertically symmetric with respect to an imaginary center line of the section. |
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claims | 1. A method of using a computer processor for automated independent technical review, the method comprising:receiving an assay result of a radioactive waste container;generating a review template;determining whether said assay result is within a predetermined parameter based on said generating said review template, said determining whether said assay result is within a predetermined parameter including comparing a total plutonium mass result to a qualification mass value, the comparing including:comparing said total plutonium mass result to a low qualification mass value:determining that said assay result is not within said predetermined parameter if said total plutonium mass result is less than said low qualification mass;comparing said total plutonium mass result to a high qualification mass value; anddetermining that said assay result is not within said predetermined parameter if said total plutonium mass result is greater than said high qualification mass value;determining whether a review is required if said assay result is not within said predetermined parameter;rejecting said assay result if said review is not required and said assay result is not within said predetermined parameter; andgenerating a report indicating that an expert review is required if an item description code for said radioactive waste container is not found within said assay result. 2. The method of claim 1, wherein said assay result is a gamma radiation assay result. 3. The method of claim 1, further including generating a comment template if said review is required. 4. The method of claim 1, wherein said generating said review template includes:generating an assay result data field including said assay result;generating a requirements field including said predetermined parameter;generating a review field including a first instruction based on said determining whether said review is required; andgenerating a rejection field including a second instruction based on said determining whether said review is required. 5. The method of claim 1, further including:determining the identity of a material in said radioactive waste container; anddetermining whether said assay result is acceptable based on said identity of said material. 6. The method of claim 1, wherein said determining whether said assay result is within said predetermined parameter includes determining whether a relative error for a plutonium isotope is within said predetermined parameter. 7. The method of claim 6, wherein said determining whether a relative error for a plutonium isotope is within said predetermined parameter includes:determining an absolute 3-sigma error for said plutonium isotope;determining a range for the weight percent of said plutonium isotope based on said absolute 3-sigma error; anddetermining that said assay result is not within said predetermined parameter if an accepted weapons grade weight percent is not within said range. 8. The method of claim 6, further including using a default isotopic if no measurement for said plutonium isotope is available. 9. The method of claim 6, wherein said determining whether said relative error for said plutonium isotope is within said predetermined parameter includes using a default isotopic parameter if said relative error is greater than about 70 percent. 10. The method of claim 9, wherein said relative error is based on a plutonium isotope for Pu240. 11. The method of claim 1, wherein said determining whether said assay result is within said predetermined parameter further includes determining whether the density of said radioactive waste container is within said predetermined parameter. 12. The method of claim 11, wherein said determining whether said density of said radioactive waste container is within said predetermined parameter includes determining that said assay result is not within said predetermined parameter if said density is greater than about 2.5 grams per cubic centimeter. 13. The method of claim 11, wherein said determining whether said density of said radioactive waste container is within said predetermined parameter includes determining that said assay result is not within said predetermined parameter if said density is less than a preselected limit. 14. The method of claim 1, wherein said determining whether said assay result is within said predetermined parameter further includes determining whether a radioactive material in said radioactive waste container is lumped. 15. The method of claim 14, wherein said determining whether said radioactive material in said radioactive waste container is lumped includes comparing the mass ratio of two gamma energies. 16. The method of claim 15, wherein said comparing the ratio of said two gamma energies includes:determining the mass ratio of a 413.71 keV gamma energy to a 129.294 keV gamma energy;determining that said assay result is not within said predetermined parameter if said ratio is greater than about 2.5. 17. The method of claim 1, wherein said determining whether said assay result is within said predetermined parameter further includes determining that said assay result is not within said predetermined parameter if a total plutonium weight percent is greater than about 10 percent. 18. The method of claim 1, wherein said determining whether said assay result is within said predetermined parameter further includes determining that said assay result is not within said predetermined parameter if a criticality safety value is greater than about 220 grams. 19. The method of claim 1, wherein said determining whether said assay result is within said predetermined parameter further includes determining that said assay result is not within said predetermined parameter if a fissile gram equivalent at 2 sigma is greater than about 220 grams. 20. The method of claim 1, wherein said determining whether said assay result is within said predetermined parameter further includes using a nuclide total result to compare a mass ratio of a first isotope and a second isotope. 21. The method of claim 20 wherein said first isotope is Pu239 and said second isotope is Am241. 22. The method of claim 21 wherein said using a nuclide total result to compare said mass ratio of a first isotope and a second isotope includes determining that said assay result is not within said predetermined parameter if said mass ratio is less than about 200. 23. The method of claim 20 wherein said first isotope is Pu239 and said second isotope is Np237. 24. The method of claim 23 wherein said using a nuclide total result to compare said mass ratio of a first isotope and a second isotope includes determining that said assay result is not within said predetermined parameter if said mass ratio is less than about 125. 25. The method of claim 1, wherein said determining whether said assay result is within said predetermined parameter further includes determining a nuclide totals result for an isotope. 26. The method of claim 25, wherein said isotope is Np237. 27. The method of claim 26, wherein said determining said nuclide totals result is not performed for said isotope if the presence of said isotope is confirmed. 28. The method of claim 25, wherein said isotope is U235. 29. The method of claim 28, wherein said determining said nuclide totals result is not performed for said isotope if the presence of said isotope is confirmed. 30. The method of claim 25, wherein said isotope is U233. 31. The method of claim 25, wherein said isotope is U238. 32. The method of claim 25, wherein said determining whether said assay result is within said predetermined parameter further includes determining that said assay result is not within said predetermined parameter if a count rate corresponding to said isotope is greater than about 5 times an error value. 33. The method of claim 1, wherein said determining whether said assay result is within said predetermined parameter further includes determining that said assay result is not within said predetermined parameter if a 400 keV transmission source peak intensity is less than about 1 percent of a calibrated intensity. 34. The method of claim 1, wherein said determining whether said assay result is within said predetermined parameter further includes:defining a segment of said radioactive waste container;determining whether a transmission source peak for said segment of said radioactive waste container is a low transmission source peak having an energy of less than about 400 keV; anddetermining that said assay result is within said predetermined parameter if said low transmission source peak is greater than about 0.1 percent of a calibrated intensity. 35. The method of claim 1, wherein said determining whether said assay result is within said predetermined parameter further includes:detecting the presence of a pulser peak;determining that said assay result is not within said predetermined parameter if said pulser peak is not detected; anddetermining that said assay result is not within said predetermined parameter if a total number of counts in said pulser peak is less than a preset fraction of an initial count rate. 36. The method of claim 1, wherein said determining whether said assay result is within said predetermined parameter further includes:detecting the presence of a reference source peak;determining that said assay result is not within said predetermined parameter if said reference source peak is not detected; anddetermining that said assay result is not within said predetermined parameter if a total number of counts in said reference source peak is less than about 50 percent of a calibrated rate. 37. The method of claim 1, wherein said determining whether said assay result is within said predetermined parameter further includes:defining a segment of said radioactive waste container;determining a live time result for said segment;determining a real time result for said segment; anddetermining that said assay result is not within said predetermined parameter if said live time result divided by said real time result is less than about 0.3. 38. The method of claim 1, wherein said determining whether said assay result is within said predetermined parameter further includes:defining a first segment and a second segment of said radioactive waste container;detecting a first radioactivity level of said first segment;detecting a second radioactivity level of said second segment;detecting a total radioactivity level of said radioactive waste container; anddetermining that said assay result is not within said predetermined parameter if said first radioactivity level and said second radioactivity level combined is greater than about 50 percent of said total radioactivity level. 39. The method of claim 38, wherein said first segment is at a bottom end of said radioactive waste container. 40. The method of claim 39, wherein said first segment is disposed against said second segment. 41. A system for automated independent technical review, the system comprising:a host system for receiving an assay result of a radioactive waste container, generating a review template, determining whether said assay result is within a predetermined parameter based on said generating said review template, determining whether a review is required if said assay result is not within said predetermined parameter, rejecting said assay result if said review is not required and said assay result is not within said predetermined parameter, and generating a report indicating that an expert review is required if an item description code for said radioactive waste container is not found within said assay result; wherein said determining whether said assay result is within a predetermined parameter includes comparing a total plutonium mass result to a qualification mass value, the comparing including: comparing said total plutonium mass result to a low qualification mass value; determining that said assay result is not within said predetermined parameter if said total plutonium mass result is less than said low qualification mass; comparing said total plutonium mass result to a high qualification mass value; and determining that said assay result is not within said predetermined parameter if said total plutonium mass result is greater than said high qualification mass value;a network coupled to said host system; anda database coupled to said host system for storing data relating to said automated independent technical review. 42. The system of claim 41, further including:a user system coupled to said network; andsaid user system accessing said host system via said network. 43. A computer-readable storage medium encoded with machine-readable computer program code for automated independent technical review, the storage medium including instructions for causing a processor to implement a method comprising:receiving an assay result of a radioactive waste container;generating a review template;determining whether said assay result is within a predetermined parameter based on said generating said review template, said determining whether said assay result is within a predetermined parameter including comparing a total plutonium mass result to a qualification mass value, the comparing including:comparing said total plutonium mass result to a low qualification mass value;determining that said assay result is not within said predetermined parameter if said total plutonium mass result is less than said low qualification mass;comparing said total plutonium mass result to a high qualification mass value; anddetermining that said assay result is not within said predetermined parameter if said total plutonium mass result is greater than said high qualification mass value;determining whether a review is required if said assay result is not within said predetermined parameter;rejecting said assay result if said review is not required and said assay result is not within said predetermined parameter; andgenerating a report indicating that an expert review is required if an item description code for said radioactive waste container is not found within said assay result. |
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claims | 1. A charged particle lithography system for transferring a pattern onto the surface of a target, comprising:a beam generator for generating a plurality of charged particle beamlets, the plurality of beamlets defining a column;a beam stop array having a surface for blocking beamlets from reaching the target surface and an array of apertures in the surface for allowing the beamlets to reach the target surface; anda modulation device for modulating the beamlets to prevent one or more of the beamlets from reaching the target surface or allow one or more of the beamlets to reach the target surface, by deflecting or not deflecting the beamlets so that the beamlets are blocked or not blocked by the beam stop array, the modulation device comprising:a substrate provided with a plurality of apertures arranged in arrays for letting the beamlets pass through the modulation device; anda plurality of modulators arranged in arrays, each modulator provided with electrodes extending on opposing sides of an aperture for generating a voltage difference across the aperture;wherein the modulators are arranged in groups, each group of modulators for directing a group of beamlets towards a single aperture in the beam stop array;wherein individual modulators within each group of modulators have an orientation such that, during generation of a voltage difference across the aperture of the individual modulator for blocking a beamlet onto the blocking surface of the beam stop array, a passing beamlet is directed to a blocking position onto the beam stop array; andwherein the blocking positions of beamlets of the group of beamlets are substantially homogeneously spread around the corresponding single aperture in the beam stop array. 2. The system according to claim 1, wherein each group of modulators is arranged to converge a corresponding group of beamlets at a common point. 3. The system according to claim 2, wherein the individual modulators of each group of modulators are rotated for deflection of the beamlets of a group of beamlets along radial lines extending from the point of convergence of the group of beamlets. 4. The system according to claim 1, wherein each group of modulators is arranged in a rectangular array. 5. The system according to claim 1, wherein each group of modulators is arranged in a radial arrangement around a centrally located axis of the corresponding groups of beamlets. 6. The system according to claim 1, wherein the modulation device includes a plurality of memory elements, each memory element storing a signal for control of one of the modulators. 7. The system according to claim 1, wherein the modulators are arranged in a two dimensional array, and wherein the rows and columns are addressed by bit-lines and word-lines. 8. The system according to claim 1, wherein each group of modulators has a center point, a virtual line substantially perpendicular to the substrate surface of the modulation device through the center point being defined as optical axis; andwherein individual modulators within each group of modulators have an orientation such that, during generation of the voltage difference across the aperture of the individual modulator, a passing beamlet is deflected in a direction along a radial line extending from the optical axis. 9. The system according to claim 8, wherein each group of modulators is arranged to converge a corresponding group of beamlets at a common point, wherein the common point of convergence for the corresponding group of beamlets is on the optical axis of the corresponding group of beamlets. 10. The system according to claim 1, wherein the modulators comprise electrodes having a concave shape. 11. A modulation device for use in a charged particle lithography system for patterning a plurality of charged particle beamlets in accordance with a pattern, the beamlets defining a column, the modulation device serving to modulate the beamlets to prevent one or more of the beamlets from reaching the target surface or allow one or more of the beamlets to reach the target surface, by deflecting or not deflecting the beamlets, the modulation device comprising:a substrate provided with a plurality of apertures arranged in arrays for letting the beamlets pass through the modulation device;a plurality of modulators arranged in arrays, each modulator provided with electrodes extending on opposing sides of an aperture for generating a voltage difference across the aperture;wherein the modulators are arranged in groups, each group of modulators for deflecting or not deflecting a group of beamlets;wherein each group of modulators has a center point, a virtual line substantially perpendicular to the substrate surface of the modulation device through the center point being defined as optical axis; andwherein individual modulators within each group of modulators have an orientation such that, during generation of the voltage difference across the aperture of the individual modulator, a passing beamlet is deflected in a direction along a radial line extending from the optical axis. 12. The device according to claim 11, wherein the modulators comprise electrodes having a concave shape. |
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047818859 | summary | The invention relates to a nuclear reactor fuel assembly including an elongated fuel channel with a square cross section and channel walls, the fuel channel having an imaginary lattice disposed therein with a box-shaped cross section having mesh openings and sides parallel to the channel walls, fuel rods containing nuclear fuel being mutually spaced apart in the mesh openings, and a water pipe spaced apart from the fuel channel by a given spacing and having a cross section spanning more than one of the mesh openings of the imaginary lattice, the given spacing being completely filled with the mesh openings of the imaginary lattice and the fuel rods disposed therein. A nuclear reactor fuel assembly of this type is known from U.S. Pat. No. 3,808,098. The known nuclear reactor fuel assembly is intended for a boiling water reactor. The water pipe thereof is a hollow cylinder and is not centrally disposed in the imaginary lattice. The water pipe forms an internal channel in the fuel assembly, through which water flows that does not boil over the entire length of the fuel assembly in the boiling water reactor. As a result, the water in the water pipe causes a better distribution of the flux of thermal neutrons and thus brings about improved reactivity in the reactor core of the boiling water reactor. In the prior art nuclear reactor fuel assembly, some subsidiary flow channels in the longitudinal direction of the fuel assembly have a different cross section from other subsidiary flow channels, which are formed only by the interspace between adjoining fuel rods. Such flow channels are formed by the intermediate space between the water pipe and adjoining fuel rods and in a boiling water reactor, a water and steam mixture flows through them in the longitudinal direction of the fuel assembly. The difference in the cross sections becomes greater as the size of the cross section of the water pipe and the the number of mesh openings or apertures of the imaginary lattice or grid encompassed by this cross section increases. Because of this difference in cross sections, the water-steam mixture flowing through the nuclear reactor fuel assembly is not uniformly distributed, which results in uneven cooling of the fuel rods of the fuel assembly. It is accordingly an object of the invention to provide a nuclear reactor fuel assembly which overcomes the hereinafore-mentioned disadvantages of the heretofore-known devices of this general type and to make the cooling of the fuel rods uniform. With the foregoing and other objects in view there is provided, in accordance with the invention, a nuclear reactor fuel assembly, comprising an elongated fuel channel with a square cross section and channel walls, the fuel channel having an imaginary lattice disposed therein with a box-shaped cross section having mesh openings and sides parallel to the channel walls, fuel rods containing nuclear fuel being mutually spaced apart in the mesh openings, and a prismatic water pipe spaced apart from the fuel channel by a given spacing and having a cross section spanning more than one of the mesh openings of the imaginary lattice, the given spacing being completely filled with the mesh openings of the imaginary lattice and the fuel rods disposed therein. The cross-sectional outline of the prismatically constructed water pipe which is defined by straight lines, makes it possible to dimension the cross section of subsidiary flow channels which are partly defined by the water pipe in such a way that a largely uniform cooling of all of the fuel rods in the nuclear reactor fuel assembly is attained. Moreover, the prismatically constructed water pipe is capable of effecting a more uniform distribution of water, which acts as the moderator substance, and nuclear fuel, and thus of increasing the reactivity of the boiling water reactor. Pages 4-7 of the ASEA Journal, 3/84, do disclose a water pipe disposed in the center of a nuclear reactor fuel assembly for a boiling water reactor, which is a hollow body having a cross-shaped cross section. However, this hollow body is not spaced apart from the elongated fuel channel; instead, its longitudinal edges are welded directly to the inside of the fuel channel. As a result, although relatively uniform cooling of the fuel rods in the fuel assembly is attained, nuclear fuel and water are nevertheless unevenly distributed in the reactor core of the boiling water reactor, since a majority of the water in the water pipe having the cross-shaped cross section is located on the inner surfaces of the fuel channel, while in the reactor core of the boiling water reactor, water is already present as the moderator substance, on the outer surfaces of the fuel channel. The nuclear reactor fuel assembly according to the invention not only provides uniform cooling of the fuel rods and increased reactivity in the boiling water reactor, but it also has the further advantage of permitting the fuel assembly components, such as spacers, fuel assembly top fitting and fuel assembly base to be adapted to the water pipe without special expense. No adaptation of the fuel channel, of any kind whatsoever, is needed. In accordance with another feature of the invention, the water pipe is centrally disposed in the imaginary lattice. In accordance with a further feature of the invention, the cross section of the water pipe is square. In accordance with an added feature of the invention, the cross section of the fuel channel has cross-sectional sides, the cross section of the water pipe is square and has cross-sectional sides, the cross-sectional sides of the water pipe are each parallel to a respective cross-sectional side of the fuel channel, and the cross-sectional sides of the water pipe are each spaced apart from a respective one of the cross-sectional sides of the fuel channel by the same distance. In accordance with a concomitant feature of the invention, the imaginary lattice (10, 11) has n.times.n mesh openings, where n 8, the cross section of the central water pipe spanning (n-6).times.(n-6) mesh openings leaving remaining mesh openings between the water pipe and the fuel channel being occupied solely by fuel rods. These advantageous embodiments of the nuclear reactor fuel assembly provide better symmetry in the distribution of the moderator substance in the form of water and of the fuel in the reactor core of the boiling water reactor and therefore permit more uniform distribution of the enrichment with fissionable isotopes in the fuel rods of the nuclear reactor fuel assembly. Other features which are considered as characteristic for the invention are set forth in the appended claims. Although the invention is illustrated and described herein as embodied in a nuclear reactor fuel assembly, it is nevertheless not intended to be limited to the details shown, since various modifications and structural changes may be made therein without departing from the spirit of the invention and within the scope and range of equivalents of the claims. |
045490830 | description | DESCRIPTION OF THE PREFERRED EMBODIMENT A phosphor screen in a cathode ray tube consists of tiny phosphor crystals having a size of around 5 to 10 micrometers. The crystals emit cathodoluminescence when an electron beam from the tube cathode has penetrated into the phosphor crystals and as a consequence thereof, a part of the energy from the penetrated electrons is converted to the energy of emitted photons (i.e. cathodoluminescence). Therefore, to understand cathodoluminescence from the phosphor screen, a comprehension of the mechanisms involved in the energy loss of the penetrating electrons in the phosphor crystals is essential. When a crystal is irradiated by an electron beam, a large part of the incident electrons penetrate into the crystal, and the residual electrons are ejected from the crystal surface (e.g. -backscattered primary electrons which are due to the elastic scattering with the lattice ions which are arranged in the surface layers of the crystal). The penetrating electrons lose their energy by elastic and inelastic collisions with the lattice ions along the electron trajectories, generating X-rays, Auger electrons, secondary electrons, electron-hole pairs, and phonons. The secondary electrons and scattered incident electrons form the electron gas plasma in the crystal and the lattice ions are excited by the electron-plasmon interaction to produce other secondary electrons having an energy which is smaller than that of the interaction electrons. Therefore, under the irradiation of the electron beam from the tube cathode on the crystal, the electrons in the ground states of the lattice ions may be directly excited into their higher energy states (excited states) by the collision with the incident electrons and the internally generated secondary electrons. The electrons in the excited states return to their ground states by emitting photons (i.e. -a radiative transition) and/or phonons (i.e. -a nonradiative transition). Phosphor crystals contain radiative transition centers, which are referred to as "activators", so that the phosphor crystals emit cathodoluminescence under the irradiation of an electron beam. The activators in some crystals are also indirectly excited through the recombination of electrons and holes generated in the crystal by the penetrated electrons. Under the given condition of the irradiation of electrons on the phosphor crystals, the number of direct excitations of the activators in the penetration volume is negligibly small as compared to the number of electrons and holes generated in the crystals. It follows that the indirect excitation of activators results in a brilliant cathodoluminescence intensity and the direct excitation of the activators results in a weak cathodoluminescence intensity. The phosphor crystals containing activators of the recombination of electrons and holes are widely used in practical cathode ray picture tubes to obtain a brilliant cathodoluminescence. Hence, the brightness of the cathodoluminescence from the phosphor screen in a practical cathode ray tube is proportional to the number of the recombinations of electrons and holes which are generated in the crystal by the irradiation of an electron beam from the tube cathode. The number n of the pairs of electrons and holes generated in the crystal by the irradiation of the electron beam is given by: EQU n=kW (1) where W is the energy dumped in the crystal and k is constant. W is given by the product of the accelerating voltage V of the electrons from the tube cathode and the electron beam density on the phosphor crystal. If the beam density on the phosphor crystal is constant, W (corresponding to n) is expressed by a linear function of the accelerating voltage V of the electron beam. It follows that under the condition that the electron beam density is constant, the brightness of the cathodoluminescence should be proportional to the accelerating voltage of the electron beam (i.e. -the voltage dependence curve of cathodoluminescence). It can be said, therefore, that when the electron beam has penetrated into the phosphor crystals, the intensity of cathodoluminescence shold be linear with respect to the accelerating voltage (V). A study of the voltage dependence curve of cathodoluminescence intensities was made with the phosphor screen in a demountable cathode-ray tube configuration as shown in FIG. 1. The phosphor crystals forming the screen lie on the face plate (substrate 6) which is transparent for luminescence viewing. With the phosphor screen prepared by ordinary methods, one obtains a linear dependence of cathodoluminescence intensities in the voltage range above 3 kV and, in the voltage range below 3 kV, the faint cathodoluminescence intensities, which deviate from the linear relationship, are obtained. The voltage of the extrapolated point of the intersection of the linear dependence curve to the abscissa of the voltage dependence curve (about 2 kV) is referred to as the "dead voltage". The dead voltage is not dependent on the type of phosphors or on the degree of chemical etching of the crystals. After extensive and careful study has been carried out on the voltage dependence curves of cathodoluminescence, it has been found that the dead voltage at 2 kV is not caused by the characteristic properties of the phosphor crystal itself but originates from foreign materials which are affixed to the surface of the phosphor crystals or from a thin film of the foreign materials which cover the phosphor crystals. In the voltage range below 2 kV, the characteristic properties of the phosphor crystals are concealed in the properties of the foreign materials on the surface of phosphor crystals, which are induced by the irradiation of the electron beam. The foreign materials affixed to or covering the surface of the phosphor crystals come from (a) the binder used in the screening process of applying the phosphor crystals to the cathode ray picture tube, (b) the materials deliberately coated on the crystals to improve the screen quality, and (c) the chemical contamination of the surface in a few atomic layers during the crystallization of the crystals, or by the exposure of the crystals in an air atmosphere during the production of the tube. When the foreign materials have been removed from the surface of the phosphor crystals, the intensities of the cathodoluminescence are linear with respect to the accelerating voltage in the voltage range below 3 kV, and the threshold voltage of the cathodoluminescence falls in the range of around a few hundred volts, depending on the phosphor crystals. Before explaining the low threshold voltage of phosphor crystals which have a clean surface, the effects of the foreign materials induced by the irradiation of the electrons on the concealment of the characteristic voltage dependence curve of the phosphor crystals must be clarified in order to more easily understand the present invention. The concealing mechanisms by the foreign materials on the surface of phosphor crystals can be explained as follows: As already described, the penetrating electrons lose their energy by elastic and inelastic collisions with the ions in the materials, with single and/or multiple scattering models, along the electron trajectories, generating X-rays, Auger electrons, secondary electrons, electron-hole pairs, and phonons. The ions in the materials are excited not only by the incident electrons but also by the internally generated secondary electrons. It is known that the internally generated secondary electrons and scattered incident primary electrons form an electron gas plasma in the crystal, and the lattice ions are also excited by the electron-plasmon interaction to produce other secondary electrons having an energy which is smaller than that of the interaction electron. The mean free path of the plasma electrons (i.e. -the average distance which an electron travels in the crystal without a collision) depends on the energy of the plasma electrons, and can be theoretically calculated from an equation. In an ordinary cathode ray tube, the mean-free path is about 10 .ANG., which is equivalent to three to five crystal lattice distances, depending on the materials used. Hence, only the secondary electrons generated in the surface volume at a depth which is smaller than the mean-free path can escape from the crystal surface, giving rise to secondary electrons which can be collected in front of the crystal. The secondary electrons which are collected in front of the crystal surface have an energy which is smaller than 50 electron volts, and are referred to as "true secondary electrons" as distinguished from the secondary electrons which cannot escape from the crystal and from the backscattered primary electrons. Each collision of a plasma electron with an ion produces one secondary electron. The probability that one penetrating electron collides with the lattice can be calculated by a Monte Carlo technique. The calculation shows that an incident electron, including a secondary electron generated in the crystal, collides a few times with the ions in a surface volume which is shallower than the mean-free path (i.e. -one entered electron may produce a few true secondary electrons). This means that the ratio (.delta.) of the true secondary electrons to the entered electrons is always greater than one when the incident electron has penetrated into the crystal. The reported experimental .delta.-values for a given crystal are nearly constant with respect to the accelerating voltage, thereby supporting the model described above. Thus, the hypothesis that the .delta.-values of the true secondary electrons are smaller than one in some voltage range, which has traditionally been believed, is incorrect. The ejection of true secondary electrons from the material leaves holes in the surface volume of the material, and the number of holes left is in accordance with the number of true secondary electrons ejected. Therefore, when the incident electrons have penetrated into the material, the material holds the positive charges (holes) in the surface volume on the irradiation side, with the result that more electrons (i.e. -true secondary electrons) are ejected from the material than electrons which have entered (.delta.>1). The positive field produced by the holes may extend outside the material (in a vacuum) and attracts the true secondary electrons and backscattered electrons. If the electrons have insufficient energy to re-enter the material, the electrons may be bound at a short distance from the material surface by an electrostatic force, and the bound electrons do not move from the surface of the material. These electrons which are fixed in front of the material are called as "surface-bound-electrons". The surface-bound-electrons are instantly formed in front of the material when the incident electrons have entered. If the irradiated material is observed from the gun side, the material is apparently covered with a negatively charged electron cloud (i.e.- the surface-bound-electrons). If one merely pays attention to the surface-bound-electrons forming in front of the material, regardless of the binding force, the surface-bound-electrons will be observed as an "electron cloud" formed in front of the material. In reality, the surface-bound-electrons always need the binding pairs that are the holes in the surface volume of the material, and they are tightly bound each other against the material surface (i.e.- the boundary of the material and a vacuum). Since the foreign material is uniformly distributed on the surface of the phosphor crystals, the negative field produced by the surface-bound-electrons on the foreign material may be extended over the phosphor crystals and effectively shields the phosphor crystals. The negative shield prevents the low energy electron beam from reaching the phosphor crystals. Consequently, no luminescence would be observed with the low energy electron beam after the very short time during which the first incident electrons have penetrated into the foreign material. Electrons which have an energy which is large enough to penetrate through the shield reach the phosphor crystals and enter into the crystals, subsequently producing the brilliant cathodoluminescence. This gives rise to a constant threshold voltage (e.g.- 2 kV) for the penetration through the shield. The real voltage dependence curve of phosphors is thus concealed in the shielding by the surface-bound-electrons outside the foreign material covering the phosphor crystals in the range below 3 kV, and the nondependence of the threshold voltage on the type of phosphors, the etching of the crystals, and the concentrations of the binder can be explained by the concealment in the shield by the surface-bound-electrons on the foreign material. With a macroscopic picture, the repulsion of the incident electrons from the phosphor crystals is a well known phenomena, and this has been interpreted for a long time as the charge-up of the phosphor crystals itself due to the smaller ratio of the true secondary electron emission. As already described above, this statement is incorrect. The crystal itself is never charged-up negatively under the irradiation of an electron beam. Rather, it is due to the surface-bound-electron, i.e.- the apparent electron cloud, formed in front of the foreign material. To observe the proper voltage dependence curve of cathodoluminescence from the phosphor screen, the phosphor crystals must have clean surfaces, and the contamination of the surfaces should be avoided in the preparation of phosphor crystals and in the tube fabrication. Since the phosphor crystal itself is a good insulator (e.g. having a resistivity which is greater than 10.sup.10 ohm-cm) and also has surface-bound-electrons under the irradiation of an electron beam, a question arises as to why the phosphor crystals which have a clean surface exhibit a different voltage dependence curve from that of the contaminated phosphor crystals. This evidence shows how the surface-bound-electron formed on the phosphor crystals are removed from the surface of the phosphor crystal. There are two possible ways to remove the surface-bound-electrons from the surface of the crystals. One possible way is the application of a positive field over the crystals. Since the surface-bound-electrons are very tightly bound in front of the crystals, the application of the positive field produced by the collecting electrodes is not large enough to remove the surface-bound-electrons. This has already been proven by the fact that the surface-bound-electrons formed on the surface cannot be removed by the field produced by the collecting electrodes. Another possible way is the elimination of the holes, which are binding pairs of the surface-bound-electrons from the surface volume. The surface-bound-electrons which lose the binding pairs (holes) are released from the surface of the crystal. The removal (or elimination) of the holes from the surface volume is easily achieved in the conductive material, which is connected to an external power source. When the irradiation of an electron beam has been made on the conductive material, the true secondary electrons are emitted from the surface of the conductors, leaving the holes in the surface volume of the conductor. The true secondary electrons may be instantly attracted by the holes in the surface volume of the conductor to form the surface-bound-electrons. When the conductor is connected to an external power source, the electons are injected into the conductor from the power source through the connected electrode. The injected electrons which have a high mobility in the conductor, migrate into the conductive material and meet with the holes. Then, the electrons recombine with the holes in the surface volume of the conductor, thereby eliminating the holes in the surface volume of the conductor. The surface-bound-electrons formed in front of the conductor, in the next moment, lose the binding pairs, and the electrons released from the surface are easily collected by the collecting electrodes. If the conductive material is disconnected from the power supply, the electrons are not injected into the conductor. Therefore, the holes in the surface volume remain in their generated places. Hence, the surface-bound-electrons may stay in front of the surface of the conductor and prevent the late arriving electrons from reaching the conductor. The repulsion of the incident electrons from the disconnected conductors is sometimes perceived as the charge up of the conductor. The phosphor crystals are tiny crystals and each of these do not possess electrodes on the surface so as to make an ohmic contact with the conductive substrate. This means that the crystals in the phosphor screen are electrically isolated from each other and are also electrically disconnected, (i.e.- floating) from the conductive substrate. Therefore, there is no way to inject the electrons into the crystals from the conductive substrate or into the conductive substrate from the crystals. Therefore, the holes in the surface volume of the phosphor crystal cannot be eliminated by means of the injection of the carriers from the conductive substrate. It has been found that if the phosphor crystals contain the recombination centers of electron-hole pairs, then the holes in the surface volume of the phosphor crystals are partially and/or totally eliminated from the surface volume when the electron-hole pairs are densely generated in the crystals. This is because the holes in the surface volume move in the crystal bulk in which the holes are recombined with the electrons at the activators during the high conductivity of the crystal. The phosphor crystal is usually an insulator, but becomes a highly conductive material during the time when the crystal contains a large amount of carriers. The conductivity of the phosphor crystals is proportional to the amount of the carriers (i.e.- electrons and holes) generated in the crystals, even though the individual carrier has a low mobility in the crystals. The conductivity is increased with an increase in the density of the carriers in the crystals. A high density of carriers is generated in the phosphor crystals under the irradiation of X-rays, electron beams and photons having an energy which is greater than the band gap of the phosphor crystals. When the phosphor crystals are conductive, the holes in the surface volume of the crystals may migrate into the bulk of the crystals. Ultimately, the holes are recombined with the electrons at the recombination centers of electron-hole pairs. Thus, the recombination of electrons and holes eliminates the carriers in the crystals. The density of the carriers generated in the crystal is decreased in time with termination of the irradiation of X-rays, electron beams or photons on the crystal, and thus the crystal ultimately becomes an insulator (i.e.- a low conductivity material). The important effect of the holes moving into the bulk from the surface volume is that a surface-bound-electron instantaneously loses its intimate binding pair (hole) and moves along the crystal surface to find another binding pair. If the electric field produced by the collecting electrodes are applied over the phosphor crystals, the surface-bound-electron which instantaneously loses the intimate binding pair may be removed from the surface of the phosphor crystal. There are two kinds of recombination centers formed in the phosphor crystals; radiative recombination centers (i.e.- activators), and non-radiative recombination centers. In practical phosphors, the number of non-radiative recombination centers is minimized. It has been found that the recombination centers of electron-hole pairs form at the impurities (or crystal defects) which can change the valences, and the recombination process of electron hole pairs at the recombination centers is triggered with the first trapped carriers, either electrons or holes, depending on the recombination centers. The recombinations centers which are triggered with the capture of either an electron or hole effectively remove the holes in the surface volume of the phosphor crystals. The surface-bound-electrons on the phosphor crystals which are arranged on the screen in an ordinary cathode ray tube are not completely removed from the surface of the phosphor crystals, and some surface-bound-electrons still remain on the surface of the phosphor crystals. To completely remove the surface-bound-electrons from the phosphor crystals, it is necessary that the phosphor crystals are screened in a special cathode ray tube having the configuration shown in FIG. 1. Using this tube configuration, a good result is obtained with respect to removing the surface-bound-electrons with a screen thickness of between two and five layers (average) of the phosphor crystals on the optically transparent and electrically conductive face plate. With thicker screens of more than 5 layers, the surface-bound-electrons partially remain on the phosphor crystals, and the amount of the remaining surface-bound-electrons is progressively increased with an increase in the number of layers of phosphor crystals. When the screen is thicker than 1 mm, the surface-bound-electrons are no longer removed from the screen. It has been found that if a metal mesh which has an electrical potential which is equal to that of the conductive face plate is placed in front of the phosphor screen or a part of the bottom of the metal mesh is embedded in the phosphor screen, the surface-bound-electrons are always completely removed from the surface of the phosphor crystals, even with the thicker screen layers. However, the transmitted luminescence intensities from the phosphor screen is decreased with increase in the screen thickness and the optimum thickness always lies between 2 and 5 layers. If the measurements of the voltage dependence curve of the cathodoluminescence intensities are made with the phosphor screen in the cathode ray tube shown in FIG. 1, the voltage dependence curve of the phosphor crystals having the clean surface exhibits the hysteresis effect as shown in FIG. 2 in the voltage range just above the threshold after a negative field (e.g. -250 volt) has been applied across the phosphor crystals. The hysteresis is observed only with the following procedure: after a negative potential has been applied to the conductive film 6, the potential of the conductive film 6 is gradually increased from zero with respect to the cathode potential. No luminescence is observed below V.sub.T in FIG. 2. At a further increase in the voltage beyond V.sub.T, the cathodoluminescence intensity gradually increases up to V.sub.H. Above V.sub.H, the cathodoluminescence intensity suddently jumps up and reaches the linear dependence which is expressed by Eq. (1). Once the phosphor screen has been irradiated with electrons having an energy which is greater than V.sub.H, the cathodoluminescence intensities are linear with an accelerating voltage of the voltage range above V.sub.T and the hysteresis around V.sub.H is not observed. The application of the negative potential to the conductive film is an essential necessity to observe V.sub.H. Once the negative potential has been applied to the conductive film 6, the hysteresis is observed once in the voltage dependence curve of the cathodoluminescence intensities if the surface of the phosphor crystals is clean. Thus, the hysteresis is reproduceable with the application of the negative potential to the conductive film 6. It has been found that the hysteresis is caused by the persistent polarization and depolarization of the phosphor crystals. Insulators are always polarized when an electric field is applied across the crystal, and are depolarized when the electric field is removed from the crystal. The phosphor crystals used in this invention differ from regular insulators in that they hold their polarization after the removal of the electric field from the crystals, i.e.- they exhibit a persistent polarization. This remaining polarization is called a persistent polarization or a persistent internal polarization. The polarity of the persistent polarization corresponds to the polarity of the applied field across the crystal. The mechanisms involved in the persistent polarization are not clearly understood. An explanation of the persistent polarization is that the phosphor crystals usually contain trapped electrons and holes. When an electric field is applied to the crystals, the electrons and holes are released from the traps, and migrate in the crystals according to the electric field. The electrons are trapped in the deep electron traps distributing on one side of the crystal, and the holes are trapped in the deep hole traps on the other side of the crystal. The separately trapped carriers may give rise to the persistent polarization. The persistent polarization is depolarized when X-rays, electrons having an energy which is greater than V.sub.H, or photons having an energy which is greater than the band gap of the phosphor crystals are irradiated on the persistently polarized crystals, and as a result thereof, carriers are generated in the crystals, resulting in an increase in their conductivity. The persistent polarization, however, is not depolarized under irradiation of either electrons having an energy which is smaller than V.sub.H or photons having an energy which is smaller than the band gap of the crystals. This invention utilizes the hysteresis appearing in the voltage dependence curve of cathodoluminescence intensities caused by the persistent polarization and depolarization of the phosphor crystals. If the surface of the phosphor crystals is clean and if the crystals show persistent polarization, the phosphor crystals in the phosphor screen may persistently have negative charges on the gun side when a negative potential has been applied to the conductive films. The negative charges on the gun side of the phosphor crystals produce a negative field which prevents the low energy electrons from reaching the phosphor crystals. An electron beam having an energy which is large enough to penetrate through the negative field produced by the persistent polarization can penetrate into the phosphor crystals and cause the depolarization of the persistently polarized phosphor crystals if the proper amount of energy is dumped on the phosphor crystals. The persistent polarization and its polarity effect on the phosphor screen can be separately determined experimentally by measuring the surface potential of the phosphor screen after negative and positive electric potentials have been respectively applied to the phosphor screen. When a negative potential has been applied to the phosphor screen, the negative surface potential is detected on the surface of the phosphor screen, and a positive surface potential is detected when a positive potential has been applied to the phosphor screen. The depolarization can be detected by measuring the surface potential of the phosphor screen. After confirming that the phosphor screen holds a persistent polarization, X-rays or electrons having an energy which is greater than V.sub.H are irradiated on the entire area of the persistently polarized phosphor screen. No surface potential is detected on the irradiated phosphor screen. Thus, persistent polarization and depolarization of the phosphor screen are in accordance with the hysteresis which has appeared in the voltage dependence curve of the cathodoluminescence. Many cathodoluminescent phosphors exhibit a persistent polarization when an electric field has been applied, and the depolarization occurs as X-rays, or electron beams having a high energy irradiate the persistently polarized crystals, and also show a hysteresis effect in the voltage dependence curve of their cathodoluminescence intensities if the surface of the phosphor crystal in the screen is not contaminated with foreign materials. Such phosphors may be, for example, zinc sulfide phosphors activated with copper or silver and coactivated with chlorine or aluminum. A part of the zinc in the above phosphors can be replaced by cadmium sulfide phosphors. Other phosphors are: the oxysulfides and oxyhalides of yttrium, gadolinium, and lanthanum, which are activated with, at least, one of the elements from the group of cerium, terbium, europium, dysprosium, samarium and praseodymium; the oxides of yttrium, gadolirium, and lanthanum which are activated with, at least, one of the elements from the group of europium, samarium and dysprosium; the sulfides, silicates and phosphates of zinc which are activated with manganese. For application of the sulfides to the phosphor screen in cathode ray picture tube, contamination of the surface should be avoided in the tube production process. The surface of the sulfides are chemically unstable in air, especially with moisture and at elevated temperature. The surface layers of the sulfides are easily oxidized during the production of the tube, and the surface layer chemically converts to the compounds containing oxygen (i.e.- a contaminated layer) in which the recombination centers in the bulk no longer act as the recombination centers of electron hole pairs. If the surface of the sulfides in the phosphor screen is kept clean, the sulfides show a persistent polarization after the electric field has been applied across the crystals, and the polarized crystals are depolarized by exposure to X-rays or high energy electrons. Therefore, the sulfides show a hysteresis in the voltage dependence curve of their cathodoluminescence intensities. A cathode ray tube in accordance with the present invention which is able to display X-ray images essentially utilizes the hysteresis appearing in the voltage dependence curve of cathodoluminescence intensities and makes an X-ray image on the phosphor screen by the steady cathodoluminescence which is visually perceived without an irritating flicker. Such an X-ray image device also utilizes the secondary electrons from the phosphor crystals for reading the X-ray images formed on the phosphor screen. The details of the formation mechanisms of the X-ray images on the phosphor screen are first explained below, referring the device which is schematically shown in FIG. 1. The phosphor crystals which have a clean surface are screened to a thickness of around 20 micrometers on the electrically conductive layer 6 formed on the face plate 7 without a binder. For a cathode ray tube in accordance with the present invention, the face plate 7 and electrically conductive layer 6 are optically transparent for the X-ray image viewers and the face plate 7 must have small absorption coefficient with respect to X-rays so as to reduce the absorption (ideally no absorption) of the X-rays by the face plate 7. Thus, more X-rays penetrate through the face plate 7 and conductive layer 6, and reach the phosphor crystals 5 where the persistently polarized phosphor crystals are depolarized by the absorbed X-rays, eventually overcoming the hysteresis appearing in the voltage dependence curve of cathodoluminescence intensity. In such an X-ray image device, the electrons from the cathode 2 are defocused and are uniformly irradiated throughout the phosphor screen 5. Therefore, the phosphor screen emits cathodoluminescence if the phosphor crystals are not persistently polarized. The phosphor crystals 5 on the conductive film 6 have a persistent polarization with the negative polarity on the gun side, when a negative potential has been applied to the conductive film 6. Then, the phosphor crystals emit no cathodoluminescence. A potential between V.sub.T and V.sub.H, e.g.- V.sub.a, is applied to the conductive film 6. The electrons from the cathode 2 are accelerated by a potential V.sub.a which does not have enough energy to penetrate through the negative field produced by the persistent polarization, and the electrons are repulsed from the negative field. The repulsed electrons are collected by the collecting electrodes 3, and the phosphor screen has no cathodoluminescence. When the persistently polarized phosphor crystals arranged on an area of the phosphor screen 5 are exposed to X-rays, the persistently polarized crystals arranged on the exposed area are depolarized and allow the electrons from cathode 2 to reach the phosphor crystals. Ultimately, the exposed area of the phosphor screen emits cathodoluminescence of the intensity B.sub. a as shown in FIG. 2. If the phosphor screen 5 of the persistently polarized crystals is exposed to X-rays transmitted from the body or materials, the phosphor screen 5 displays the projected X-ray image of the space distribution of the transmitted X-rays by cathodoluminescence. The X-ray image is maintained on the phosphor screen 5 until a negative potential is applied to the conductive film 6, i.e. an erasing process. Then, another X-ray image can be displayed on the same phosphor screen 5 without any interference with respect to the previously displayed X-ray image. There is a threshold D with respect to the exposed dosage of X-rays needed to depolarize the persistently polarized phosphor crystals. Threshold D can be empirically, but not theoretically, determined by the following procedure. The conductive film 6 has a constant potential V.sub.a after a negative potential (e.g.-300 volts) has been applied. No luminescence or a faint luminescence is observed from the phosphor screen 5. Then, X-rays of various dosages are irradiated on phosphor screen 5 through face plate 7. FIG. 3 shows a typical curve representing the relationship between the depolarization (detected by the cathodoluminescence B.sub.a) and the dosages of X-ray exposure. If the phosphor crystals in the phosphor screen absorb a dosage of X-rays which is more than threshold D, the persistently polarized phosphor crystals are depolarized, resulting in the constant cathodoluminescence B.sub.a on the screen. If a dosage which is less than D is irradiated on the phosphor screen, the phosphor crystals remain persistently polarized and no luminescence is observed. The value of the threshold D varies with the kind of phosphors, and with screen conditions. A brighter cathodoluminescence allows a clear and high contrast X-ray image on the phosphor screen which may allow one to observe the X-ray image on the phosphor screen in a lighted room. The cathodoluminescence intensity B.sub.a of the phosphor screen is determined by the applied voltage V.sub.a of the conductive film 6; the cathodoluminescence intensity increases as the applied voltage is increased. If the applied voltage V.sub.a is near V.sub.H, a high cathodoluminescence intensity is expected, but the intensity of the faint cathodoluminescence (i.e.- the background luminescence) is also increased with an increase in V.sub.a. Consequently, the contrast of the X-ray image on the phosphor screen, (estimated from the ratio of B.sub.a to the background cathodoluminescence) becomes poor. Thus, a poor contrast X-ray image is obtained when the applied voltage is near V.sub.H. The cathodoluminescence intensity and the contrast of the X-ray image on the phosphor screen are markedly improved if a potential V.sub.F which is slightly below V.sub.H is applied to the first collecting electrode 3. FIG. 4 shows an improved hysteresis curve with the application of V.sub.F to the first collecting electrode 3. It can be seen from FIG. 4 that the concavity of the hysteresis curve does not become affected with the application of V.sub.F to the first collecting electrode 3 (i.e.- the background luminescence intensity does not change with V.sub.F) but the cathodoluminescence intensities with a downward going voltage from V.sub.H are markedly changed with the potential applied to the first collecting electrode 3 (i.e.- a convex curve in the hysteresis curve is obtained with V.sub.F). If V.sub.F is applied to the first collecting electrode 3, the phosphor screen emits almost constant luminescence intensity in the voltage range between V.sub.T and V.sub.H, when the applied voltage at the conductive film 6 is decreased from V.sub.H. If a potential greater than V.sub.H is applied to the first collecting electrode 3, a single curve which has a constant intensity between V.sub.T and V.sub.F is obtained instead of the hysteresis curve. Thus, V.sub.F (which is smaller than V.sub.H) should be applied to the collecting electrode 3 to obtain the improved hysteresis curve. Because the background cathodoluminescence intensity is more dependent upon the conditions of the phosphor screen and the potential of the conductive film 6, rather than on the potential at the collecting electrode 3, the contrast of the X-ray image on the phosphor screen obtained at a potential V.sub.a of the conductive film 6 is significantly improved (about double in the case of FIG. 4) by the application of V.sub.F to collecting electrode 3. An explanation of the improved hysteresis curve shown in FIG. 4 is that the electrons from cathode 2 are accelerated by the potential V.sub.F at the first collecting electrode 3, rather than the potential at the conductive film 6, and the electrons have an energy of eV.sub.F. The potential in front of the phosphor crystals, produced by the persistent polarization of the phosphor crystals, is eV.sub.H which is greater than an energy of eV.sub.F. Therefore, electrons having an energy of eV.sub.F cannot penetrate through the shielding field V.sub.H and the electrons which do not reach the phosphor crystals are collected by the first collecting electrode 3. When the persistently polarized crystals are depolarized, electrons having an energy of eV.sub.F reach and penetrate into the phosphor crystals, giving rise to a constant cathodoluminescence intensity between V.sub.T and V.sub.F. The threshold voltage V.sub.T is probably determined by the conditions of the phosphor screen; however, the reason for this is not yet clear. Similarly, the improved hysteresis curve is also obtained if the electrons from the cathode 2 are accelerated with the potential V.sub.F being applied to other electrodes, instead of the first collecting electrode 3, e.g.- an anode placed in front of the cathode 2 or a metal mesh placed in front of the phosphor screen 5. FIG. 5 shows a schematical diagram of a cathode ray tube which is able to display the X-ray images, and illustrated in order to explain the practical operation of the present invention. An optically transparent and electrically conductive thin film 16 of tin oxide containing indium oxide is coated on face plate 17 which is an optically transparent glass that contains no heavy elements (e.g. lead, Pb) having a large absorption coefficient of X-rays. A slurry made from the mixture of a conductive fine powder (e.g.- carbon, iron oxide and the like) and photosensitized polyvinyl alcohol is uniformly coated on the conductive film 16. The dried slurry layer is exposed under ultraviolet light through a mesh having holes with a diameter of 100 micrometers, said mesh being placed in front of the face plate 17. The exposed layer is developed with warm water to form the second collecting electrodes 14 on the conductive film 16. Then, the slurry of the phosphor crystals (e.g. gadolinium oxysulfide activated with terbium, Gd.sub.2 O.sub.2 S:Tb) and photosensitized polyvinyl alcohol is uniformly coated on the conductive film 16 having the second collecting electrodes 14, and the dried layer is exposed to the ultraviolet light from face plate 17. The exposed phosphor screen is developed with warm water. We then obtain a phosphor screen 15 resulting from the phosphor crystals lying down in the areas of the conductive film 16 on which the collecting electrodes 14 have not been formed. Face plate 17, which has the phosphor screen 15 and second collecting electrodes 14, is mounted in envelope 10 containing cathodes 11, anodes 12, and second collecting electrodes 13. When ordinary cathode ray tube fabrication processes have been carried out, the device which is schematically shown in FIG. 5 is obtained. In this device, the hysteresis curve shown in FIG. 2 is obtained between 110 and 210 volts after a negative potential (e.g. -300 volt) has been applied to the conductive and transparent film 16 for 1 millisecond, anodes 12 having 90 volts applied thereto and the first collecting electrodes 13 having 100 volts applied thereto. The improved hysteresis curve shown in FIG. 4 is obtained in the voltage range between 110 and 210 volts, if anodes 12 have 180 volts applied thereto, the first collecting electrodes 13 have 140 volts applied thereto and the conductive film 16 has been supplied with minus 300 volts for 1 millisecond. In order to demonstrate the display of X-ray images on the phosphor screen, the cathode ray tube should be set for the following conditions: the cathodes 11 are grounded, the anodes 12 have 180 volts applied thereto, the first collecting electrodes 13 have 150 volts applied thereto, the transparent and conductive film 16 has 160 volts applied thereto after -300 volts has been applied to the conductive film 16 for 1 millisecond. The phosphor screen in the device emits no luminescence or a faint luminescence under the above conditions. If X-rays irradiate the phosphor screen through face plate 17 for a moment, the exposed phosphor screen allows the reaching of the electrons from the cathodes 11 and emits cathodoluminescence. If a human body stands between the X-ray source and the X-ray image device, the latent X-ray image of the inside of the body is recorded on the phosphor screen 15, and the phosphor screen 15 continuously displays the X-ray image on the phosphor screen 15 until the latent X-ray image has been erased by the application of a negative potential to the transparent and conductive film 16. This means that the latent X-ray image recorded in the phosphor screen does not fade with time and the X-ray image device can continuously and intermittently display an X-ray image on the phosphor screen until the latent X-ray image on the phosphor screen is erased. The cathodoluminescence intensity from the phosphor screen is bright enough to observe the X-ray image on the phosphor screen in a room illuminated with regular lighting, instead of in a dark or dimly lit room. When the erasing process has been applied to the phosphor screen, the latent X-ray image is completely erased from the phosphor screen, and the phosphor screen in the X-ray image device is restored to record other X-ray images on the phosphor screen. However, the keeping and referring occasionally to the X-ray images of patients is essential to the operation of hospitals and clinics. To respond to these requirements, the X-ray images by cathodoluminescence or latent X-ray images on the phosphor screen should be output from the phosphor screen, and recorded onto recording media (e.g. a memory in computers) before the application of the erasing process of the phosphor screen. The easiest way to record the X-ray image on the phosphor screen, which are produced by the steady cathodoluminescence, may be by reading same with the image tube for one frame. A more reliable way would be the monitoring of a change in the electrical current of the first or second collecting electrodes. As already described, most of the electrons from the cathode are collected by the first collecting electrode if the phosphor crystals in the phosphor screen are persistently polarized, and the second collecting electrode does not collect (or collects a small amount of) the electrons, as schematically shown in FIG. 6. When the phosphor crystals are depolarized, the electrons from the cathode reach and penetrate into the phosphor crystals, generating and emitting the true secondary electrons with the ratio of a few true secondary electrons per one entered electron. The second collecting electrode E.sub.C2 collects an amount of true secondary electrons which is equal to the amount of electrons entering into the phosphor crystals, and the residual electrons reenter into the phosphor crystals, as schematically shown in FIG. 7, and the first collecting electrode E.sub.C1 collects no electrons or a very small amount of the electrons. Thus, in the cathode ray tube, the electrons starting from the cathode make a closed circuit at the phosphor screen via the true secondary electrons, instead of passing through the phosphor crystals as has been considered traditionally, and back to cathode through the collecting electrode and circuit outside of the tube so as to make a complete closed circuit. FIGS. 8a and 8b show the change in electrical current detected at first (i.sub.1) and second (i.sub.2) collecting electrodes, before and after X-rays expose the entire phosphor screen. However, the X-ray image on the phosphor screen cannot be detected by a simple reading of the change in the current of the collecting electrodes. To detect the X-ray image on the phosphor screen, the phosphor screen should be addressed, and the reading of the change in the current of the collecting electrodes should be synchronized with the scanning of the address of the corresponding location on the phosphor screen. The addressing on the phosphor screen can be made by the scanning of a sharply focused electron beam from a second electron gun (i.e.- a reading gun) from which the electrons are accelerated with voltage greater than V.sub.H shown in FIG. 2, but the beam power (i.e.- a product of the accelerating voltage and beam density) should be small so as to avoid the depolarization of the persistently polarized phosphor crystals. The resolution of the X-ray image taken from the phosphor screen is determined by the beam size of the electron beam from the reading gun; a small size results in a better recorded X-ray image. The size of the phosphor crystals of the phosphor screen is about 10 micrometers and this limits the highest resolution. Therefore, a good result will be obtained with the size of the electron beam from the reading gun being between 10 and 300 micrometers, and more preferably between 20 and 100 micrometers, and most preferably between 30 and 60 micrometers. The electron beam is more easily focused with a high accelerating voltage, but a higher voltage needs a good electrical insulator to prevent the breakdown of the electrical circuits. This will limit the upper voltage of the accelerating voltage of the reading gun. Therefore, the accelerating voltage of the electron beam from the reading gun is preferably higher than 500 volts but smaller than 30 kilovolts, and more preferably between 1 and 10 kilovolts, and most preferably around 5 kilovolts. There is, as already described, a threshold dosage of X-rays, electron beams or photons having an energy which is greater than the band gap in order to depolarize the persistently polarized phosphor crystals. All of the radiation types mentioned above generate electron-hole pairs in the phosphor crystals when they penetrate into the crystals. It can be more precisely said that the threshold of the depolarization is determined by the number of electron hole pairs generated in the phosphor crystals per unit time. The threshold number of the electron-hole pairs for the depolarization is theoretically unclear, and has been empirically determined. It is found that the scanning of an electron beam which has a 5 kV accelerating voltage and a 100 .mu.m beam size, on the persistently polarized phosphor screen, results in polarized crystals which are not depolarized if the electron beam current is between 0.001 and 1 .mu.A, and preferably between 0.01 and 0.7 .mu.A, and most preferably around 0.3 .mu.A. When the reading electron beam is irradiated on the phosphor screen, for displaying the X-ray image, the electric currents of the first and second collecting electrodes, E.sub.C1 and E.sub.C2 are changed with the period of irradiation of the reading electron beam on the X-ray images on the phosphor screen. When the reading electron beam is irradiated on the persistently polarized phosphor crystal, as schematically shown in FIG. 9, the electron beam may penetrate into the phosphor crystal, and the crystal emits the true secondary electrons and a faint cathodoluminescence. The emitted true secondary electrons are repulsed by the negative field produced by the persistent polarization of the phosphor crystal and most of them are collected by the first collecting electrode E.sub.C1. When the phosphor crystals are depolarized, the reading electron beam also penetrates into the phosphor crystals, emitting the faint cathodoluminescence and the true secondary electrons, as schematically shown in FIG. 10. In this case, some of the emitted secondary electrons reenter into the phosphor crystal by the attraction of the holes generated in the phosphor crystal by the emission of the true secondary electrons. The residual electrons of the true secondary electrons are collected by either the first collecting electrode E.sub.C1 or second collecting electrode E.sub.C2. Therefore, the electrical current of the first collecting electrode E.sub.C1 is decreased during the period of the scanning of the reading electron beam on the phosphor crystals which have been depolarized. If the electric current of the second collecting electrode E.sub.C2 is detected, the current is increased during the period of the scanning of the reading electron beam on the phosphor crystal which have been depolarized. Therefore, the change of the electric current of the collecting electrodes is synchronously detected during the scanning of the reading electron beam on the phosphor screen, and the X-ray image on the phosphor screen can be detected remotely from the cathode ray tube. The image information taken from the phosphor screen can be recorded and stored in ordinary memory media, such as magnetic tape, magnetic disks, optical disks, electonic memories and the like. The recorded image information can occasionally be output from the recorded media, and displayed on the phosphor screen of an ordinary cathode ray tube for diagnosis or reference purposes by doctors in hospitals and clinics, and by inspectors in quality control areas of manufacturing. As noted above, the change in the electric current of the collecting electrodes is synchronously detected during the scanning of the reading electron beam on the phosphor screen and the X-ray image displayed on the phosphor screen can be detected remotely from the cathode ray tube. FIG. 11 illustrates the wave form of the change in current at a collecting electrode when the reading electron beam scans the depolarized phosphor cyrstals. FIG. 12 illustrates a concrete example of an application of the present invention to a cathode ray tube. It is noted that elements 30 and 31 are illustrated symbolically as current meters. In fact, in actual usage these element would be current detecting devices which would be utilized to generate current signals corresponding to that illustrated in FIG. 11 of the drawing figures. In operation, the deflection coil 23 can be used to control the reading beam emitted by the reading electron gun 22 in a fashion which is exactly analogous to that of the operation of a standard commercial TV picture tube and accordingly, a detailed description of the control thereof has been omitted. Similarly, the correlation of the currents detected by elements 30 and 31 with the position of the reading electron beam generated by the electron gun 22 as controlled by the deflection coil 23 is also known to those skilled in the art and therefore has been omitted. Although the present invention has been fully described by way of examples with reference to the accompanying drawings, it is to be noted that various changes and modifications will be apparent to those skilled in the art and accordingly, unless such changes and modifications depart from the true scope of the present invention, they should be construed as being included therein. |
062460633 | abstract | In a radiation image storage panel having a stimulable phosphor layer of a stimulable phosphor, and a surface protective film, the surface protective film exhibits scattering with a scattering length of 5 to 80 .mu.m observed at a main wavelength of stimulated emission from the stimulable phosphor. |
061817627 | claims | 1. A nuclear fuel bundle comprising: a plurality of elongated, generally parallel nuclear fuel rods containing nuclear fuel and arranged in a matrix thereof with peripheral rods surrounding interior rods, each of said peripheral rods having a first peak power limit higher than a second peak power limit of said interior rods prior to a first fission chain reaction of the nuclear fuel in the bundle in a nuclear reactor for power generation, each fuel rod containing nuclear fuel, the magnitude of nuclear fuel within each of said peripheral rods being less than the magnitude of nuclear fuel within each of said interior rods, the nuclear fuel within said peripheral rods having substantially the same power output as the nuclear fuel of said interior rods, each of the peripheral rods having a gas plenum volume in excess of the gas plenum volume of each of said interior rods. a plurality of elongated, generally parallel nuclear fuel rods containing nuclear fuel and arranged in a rectilinear array thereof having edge fuel rods about the periphery of the bundle and fuel rods interior of said edge rods, said edge and interior rods having respective discrete peak power limits, the peak power limit for said edge rods being higher than said peak power limit for said interior rods prior to a first fission chain reaction of the nuclear fuel in the bundle in a nuclear reactor, each of said plurality of rods having a plurality of fuel pellets stacked one on top of the other forming a column thereof within the rod, the column of fuel pellets within each of said edge rods having a length less than the length of the column of fuel pellets within each of said interior rods, said plurality of rods having a common length, the column of fuel pellets in each edge rod forming a gas plenum volume adjacent one end thereof greater than a gas plenum volume adjacent a corresponding end of each said interior rod. a plurality of elongated, generally parallel nuclear fuel rods containing nuclear fuel and arranged in a rectilinear array thereof having edge fuel rods about the periphery of the bundle and fuel rods interior of said edge rods, said edge and interior rods having discrete peak power limits, the peak power limit for said edge rods being higher than said peak power limit for said interior rods prior to a first fission chain reaction of the nuclear fuel in the bundle in a nuclear reactor, each fuel rod containing nuclear fuel and a gas plenum within said rod, the magnitude of nuclear fuel in each said edge rod being less than the magnitude of the nuclear fuel within each said interior rod, a ratio of plenum volume to nuclear fuel volume of each said edge rod being greater than a ratio of plenum volume to nuclear fuel volume of each said interior rod. 2. A nuclear fuel bundle comprising: 3. A nuclear fuel bundle comprising: |
047330899 | claims | 1. A radiographic intensifying screen, comprising a support of a plastic film and at least one phosphor layer comprising a binder and a phosphor dispersed therein, in which said support is provided on a surface facing said phosphor layer with a great number of pits having a mean depth of 1-10 .mu.m, inclusive, a maximum depth of more than 1 .mu.m ranging to 50 .mu.m, inclusive, and a means diameter of 10-50 .mu.m, inclusive, whereby light emitted by said phosphor and advancing towards said surface of said support is reflected diffusely. 2. The radiographic intensifying screen as claimed in claim 1, in which said pits have a means depth of 1-5 .mu.m, inclusive. 3. The radiographic intensfying screen as claimed in claim 1, which said pits have a maximum depth of 2-20 .mu.m, inclusive. 4. The radiographic intensifying screen as claimed in claim 1, 2 or 3, in which said binder comprises a linear polyester as a principal component. 5. The radiographic intensifying screen as claimed in claim 1, 2 or 3 in which said binder comprises nitrocellulose as a principal component. 6. The radiographic intensifying screen as claimed in claim 1, 2 or 3, in which said binder comprises a mixture of a linear polyester and nitrocellulose as a principal component. 7. The radiographic intensifying screen as claimed in claim 1, 2 or 3, in which said pits are those formed by applying hard solid particles onto the surface of said support at high speed. 8. A process for the preparation of a radiographic intensifying screen comprising a support of a plastic film and at least one phosphor layer comprising a binder and a phosphor dispersed therein, in which said support is provided on a surface facing said phosphor layer with a great number of pits having a mean depth of 1-10 .mu.m, inclusive, a maximum depth of the than 1 .mu.m ranging to 50 .mu.m, inclusive, and a mean diameter of 10-50 .mu.m, inclusive, said process comprising applying hard solid particles onto said surface of said support at high speed to form said pits. |
abstract | A reactor core, comprising: an outermost region; a core region surrounded by said outermost region; a plurality of fuel support members, each of which is disposed at a lower end portion of said outermost region and said core region; and a plurality of fuel assemblies loaded in said outermost region and said core region and supported by said fuel support members, wherein a plurality of fuel assemblies disposed in said core region include a plurality of first fuel assemblies, each of which is inserted into a first coolant passage which is formed in said fuel support member and has a first resistor having an opening, and a plurality of second fuel assemblies, each of which is individually inserted into each of second coolant passage which is formed in said fuel support member and has a second resistor having an opening and a larger pressure loss than that of said first resistor; and, four fuel assemblies, each of which is adjacent to each of four lateral sides of each of a plurality of first fuel assemblies, include either three or four second fuel assemblies. |
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claims | 1. A device for ionizing and separating the constituents of a chemical mixture into ions having a relatively low mass to charge ratio (M 1 ) and ions having a relatively high mass to charge ratio (M 2 ), said device comprising: a wall surrounding an elongated chamber, said wall defining a longitudinal axis and having a first end and a second end, wherein said first end is formed with an opening and the chemical mixture is mounted on said wall for exposure to said chamber; a central electrode positioned in said chamber and oriented substantially along said longitudinal axis; a means for introducing a working gas into said chamber for ionization, said working gas having a mass to charge ratio (M w ) in said chamber between M 1 and M 2 (M 1 less than M w less than M 2 ); a means for generating an axially oriented magnetic field in said chamber; and a means for generating a radially oriented electric field crossed with said magnetic field and configured to cause said working gas to sputter and ionize the chemical mixture and to confine said low mass ions (M 1 ) for exit from said chamber though said first end and to direct said high mass ions (M 2 ) into said central electrode for capture thereon. 2. A device as recited in claim 1 wherein said means for generating a radially oriented electric field in said chamber comprises a voltage source connected to said wall and to said central electrode to establish a voltage difference therebetween. claim 1 3. A device as recited in claim 1 wherein said means for generating a radially oriented electric field in said chamber comprises a plurality of ring electrodes positioned in said chamber near one of said ends of said wall. claim 1 4. A device as recited in claim 1 wherein said means for generating a radially oriented electric field in said chamber comprises a voltage source connected to said wall and to said central electrode to establish a voltage difference therebetween, and a plurality of ring electrodes positioned in said chamber near one of said ends of said wall. claim 1 5. A device as recited in claim 1 wherein said means for introducing a working gas into said chamber comprises at least one hole formed in said wall. claim 1 6. A device as recited in claim 5 wherein said central electrode is formed with a gas-box for accumulation of said working gas and a means for transferring said working gas from said gas-box, outside said wall and then through said hole in said wall to reintroduce said working gas into said chamber. claim 5 7. A device as recited in claim 6 wherein said transferring means comprises a duct and said device further comprises a control valve for selectively metering said working gas through said duct. claim 6 8. A device as recited in claim 1 wherein said central electrode is formed with an outer surface facing said chamber and said outer surface is textured to minimize the loss of central electrode material from sputtering of the central electrode. claim 1 9. A device as recited in claim 1 wherein said central electrode is formed with an outer surface facing said chamber and said outer surface is formed with projections extending radially therefrom to minimize the loss of central electrode material from sputtering of the central electrode. claim 1 10. A device as recited in claim 9 wherein said projections are substantially disk shaped. claim 9 11. A device as recited in claim 1 wherein tiles are formed of said chemical mixture and said tiles are mounted on said wall. claim 1 12. A device as recited in claim 1 wherein said wall is made of said chemical mixture. claim 1 13. A device for sputtering and ionizing a chemical mixture and separating ions of said chemical mixture according to their mass to charge ratio, said device comprising: a wall surrounding a volume and formed with an opening, wherein said chemical mixture is mounted on said wall for exposure to said volume; a central electrode positioned in said volume and distanced from said wall; a means for introducing a working gas into said volume; a means for generating a magnetic field in said volume; and a means for generating an electric field in said volume directed inwardly from said wall to said central electrode, said magnetic and electric fields being configured to cause a portion of said working gas to become fast neutrals that are directed towards the chemical mixture to sputter the chemical mixture into said volume for ionization into ions in said electric and magnetic fields, said electric and magnetic fields being configured to place ions having a mass to charge ratio less than a predetermined mass to charge ratio on trajectories for exit from said volume through said opening, and to place ions having a mass to charge ratio greater than said predetermined mass to charge ratio on trajectories into said central electrode for capture thereon. 14. A device as recited in claim 13 wherein said means for introducing a working gas into said volume comprises at least one hole formed in said wall and wherein said central electrode is formed with a gas-box for accumulation of said working gas and a means for transferring said working gas from said gas-box, outside said wall, and then back into said volume through said hole in said wall. claim 13 15. A device as recited in claim 13 wherein said central electrode is formed with an outer surface facing said volume and said outer surface is formed with projections extending from said central electrode to minimize the loss of central electrode material from sputtering of the central electrode. claim 13 16. A method for separating a chemical mixture into constituents, said method comprising: providing a wall that surrounds an elongated chamber, with said chemical mixture mounted on said wall for exposure to said chamber, said wall defining a longitudinal axis and having a first end and a second end, said first end formed with an opening; generating an axially oriented magnetic field in said chamber; generating a radially oriented electric field in said chamber; aligning a central electrode along said axis in said chamber; and introducing a working gas into said chamber for interaction with said magnetic and electric fields to sputter said chemical mixture, for subsequent ionization of said chemical mixture to ions in said chamber, said electric and magnetic fields configured to confine ions having a mass to charge ratio below a predetermined mass to charge ratio for exit from said chamber through said first end and to direct ions having a mass to charge ratio above said predetermined mass to charge ratio into said central electrode for capture thereon. 17. A method as recited in claim 16 wherein said working gas comprises the gas of a noble element. claim 16 18. A method as recited in claim 16 wherein said chemical mixture comprises an alloy of Zirconium and Hafnium. claim 16 19. A method as recited in claim 18 wherein said working gas comprises Xenon. claim 18 |
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051732173 | claims | 1. A method of containing wastes and comprising, placing the wastes in a container and introducing a grout slurry into the container to fill voids in the container, thereby to embed the wastes into a solidified mass upon setting of the grout slurry, wherein the improvement comprises, closing the container with the wastes therein, piercing the container with hollow cutting means, introducing the grout slurry into the container through the cutting means, and retaining the cutting means in the container. 2. A method as claimed in claim 1, wherein the wastes are disposed within a crate, the crate is placed within the container, and the cutting means pierces the container and the crate so as to introduce the grout slurry into the crate. 3. A method as claimed in claim 2, wherein the wastes are disposed within a plurality of said crates, and a corresponding plurality of said cutting means pierce the container at a corresponding plurality of locations and pierce the respective crate. 4. A method as claimed in claim 1, wherein the wastes comprise a glove box and the contents thereof, and the cutting means pierces the glove box in the container so as to introduce the grout slurry into the glove box. 5. A method as claimed in claim 4, wherein the wastes are disposed within a plurality of said gloveboxes, and a corresponding plurality of said cutting means pierce the container at a corresponding plurality of locations and pierce the respective said glove box so as to introduce the grout slurry therein. 6. A method as claimed in claim 1, including venting the container whilst introducing the grout slurry. 7. A method as claimed in claim 6, wherein at least some of said grout slurry is transversely dispersed by the cutting means. 8. Apparatus for containing wastes and comprising a closable container for enclosing the wastes, hollow cutting means for piercing the container and for introducing a grout slurry therethrough into the container, the cutting means being detachable so as to be retained in the container after the grout slurry has been introduced therein. 9. Apparatus as claimed in claim 8, including a crate for housing the waste in the container, the cutting means being adapted to pierce the crate in the container so as to introduce the grout slurry into the crate. 10. Apparatus as claimed in claim 8, wherein the wastes comprise a glove box and the contents thereof, and the cutting means is adapted to pierce the glove box in the container so as to introduce the grout slurry into the glove box. 11. Apparatus as claimed in claim 8, wherein the container has a wall with a port means therein for locating the cutting means. 12. Apparatus as claimed in claim 11, wherein the port means includes vent means for venting air displaced from the container by the grout slurry. 13. Apparatus as claimed in claim 11, wherein the cutting means comprises a cutting tool, a stem portion arranged to drive the cutting tool, and a bearing means in which the stem portion is locatable, the bearing means being arranged to locate in the port means. 14. Apparatus as claimed in claim 13, including means for rotating the stem portion and for axially displacing the stem portion and thereby the cutting tool, and releasable means for connecting the stem portion to the rotating and axially displacing means. 15. Apparatus as claimed in claim 14, including grout filling means releasably connectable to the stem portion. 16. Apparatus as claimed in claim 14, wherein a plurality of said ports are provided for locating a corresponding plurality of respective said cutting means. 17. Apparatus as claimed in claim 14, wherein the cutting means includes means for discharging said grout slurry transversely in the container. 18. Apparatus for containing wastes and comprising, a) a container for containing watstes; b) a lid for the container; c) at least one port in the lid, the port comprising a converging well, a tube extending through the well, a bursting disc at that end of the tube inside the container, a vent in the well from the container, a locking screw extending through the wall of the tube at the other end thereof, and a disconnectable cover plate for the port; d) a cutting device comprising a hollow rotary cutting tool, a hollow stem joined at one end to the cutting tool, and a cylindrical bearing block about the stem arranged to locate in the tube and be locked therein by the locking screw; e) a connector arranged to releasably connect with the other end of the stem; f) a drive motor having a drive shaft connectable to the connector; g) rack and pinion means for displacing the motor and thereby the stem and cutting tool axially; h) a flexible feed pipe for grout slurry releasably connectable to the other end of the stem; i) and a flexible vent pipe releaseable connectable to the vent, whereby in operation with the bearing block located in the tube, rotation and axial displacement of the stem by the motor causes the cutting tool to pierce the bursting disc and enter the container, and subsequent connection of the grout slurry feed pipe to the stem without withdrawal of the cutting tool from the container causes a grout slurry to be discharged into the container and thereby embed the wastes when set, the cutting tool being embedded in the container, and the cover plate then being fitted to seal said port. 19. Apparatus as claimed in claim 18, wherein the wastes comprises a glove box and the contents thereof, and wherein a crate in which the glove box is located is inside the container, the cutting device being arranged to pierce the container, the crate, and the glove box. |
description | This is the national stage of International Application PCT/HU2011/000048, filed May 20, 2011. The inventive concept and the object of the invention are based on the teaching already known from the earlier times, according to which the utilization of nuclear reactors can be augmented with the transmutation of certain elements into other elements of higher value from certain aspects on the industrial scale (i.e. in relatively large amounts). [Related Hungarian patent application was filed under no. P 88 06077 on 28 Nov. 1988 by Péter Teleki, entitled to “Method of utilizing the (n, γ) reaction of thermal neutrons”; an international patent application was filed under no. PCT/HU89/00054 and entitled to “Method of Utilizing the (n, γ) Reaction of Thermal Neutrons”, as well as Canadian Patent no. 2003671 entitled to “Method of Utilizing the (n, γ) Reaction of Thermal Neutrons.”] The above mentioned documents disclose the transmutability of elements Yb and W, such as Yb→Lu, as well as W→Re, wherein the product obtained can be considered as an alloy of at least two components (other daughter elements, e.g. Hf and/or Os, can also form), wherein said product (and also the target) is preferably in the form of a sheet. To define the invention, the already known details are to be completed hereby in three further aspects. 1. Any lanthanide can be produced from an element located in the periodic system of elements antecedently to it; however, it is of great economical importance of the following type of target(mother element)→product(daughter element) transitions: Nd→Pm, Sm→Eu, Er→Tm, Yb→Lu. Further transitions of elements are also possible, e.g. Gd→Tb, as well as in the case of platinum metals, e.g. Ru→Rh, Re→Os (note: Re is not a platinum metal), as well as W→Re, W→Re→Os. (These transitions of elements—i.e. transmutations—are, of course, known in the literature and hence, do not form part of the invention, but are parts of the present disclosure.) 2. The physical form of the target(mother element) in the practice is not limited; e.g. powders, as well as metal lumps or granules thereof are equally preferred. It should be here noted that, in general, metal powders are inflammable and hence it is more preferred when they are provided in the form of a carbide, nitride, oxide or silicide. However, due to the large effective neutron capture cross-section of boron (B), boride variants thereof should not be made use of. Moreover, fluoride and sulphide components should be avoided due to their chemically aggressive nature, nevertheless, they are not forbidden. The above definition requires no further explanation. 3. A detailed enough disclosure of the specific cassette and/or container (from now on being distinguished from one another) suitable for the arrangement of the target is essential for a complete teaching of the invention; said means—depending on the embodiments considered—can be placed outside of the reactor shell (reactor envelope) and/or within the irradiation channel of the reactor. The object of the invention is to produce the daughter elements exemplified here on the industrial scale, and also to reach a considerable increase in productivity. A striking example is the position of Os in the global market. Osmium (Os) is the hardest metal on Earth; it is about twice as hard as tungsten (W) and can be used as an alloying element thereof, however, its amount present in global trade is less than 100 kg per year. By the inventive process, an amount of about 1000 kg per year can be produced, per reactor. Reverting now to the accomplishment of the object of the present invention, in an industrial utilization of nuclear reactors not only the so-called irradiation channels can be exploited but the targets to be irradiated can be deployed directly next to the outer casing of the reactor shell (reactor envelope) which results in a significant increase in the amount of the obtainable product. (See the problem of Os above.) This solution will affect neither the neutron balance of the reactor nor the other processes taking place within the reactor because the target is located outside of the reactor and the target (mother element)→product (daughter element) transmutation nuclear reactions are effected by the anyway harmful “waste” neutrons. Naturally, for each type of nuclear reactors there will be zones that are preferential; these zones must be selected from reactor type to reactor type. It is preferred if there is no shielding against neutron radiation between the target and the reactor envelope, said shielding has to be fully deployed behind the whole system of targets. Preferably, but not necessary, there is a neutron thermalizing moderator between the target and the reactor shell that decelerates the neutron shower to a thermal level. To this end, e.g. reactor grade graphite could be advantageously used which can be applied between e.g. aluminum sheets within the cassette already mentioned. The target can also be arranged in a further cassette; then a neutron reflector (mirror) can be arranged from the outside—also within a separate cassette—in which neutrons are scattered back towards the target. This zone is also preferred, but not a requisite, and thus its application is upon discretion. It is noted that said reflector (mirror) zone, similarly to the moderator zone, can be canned by aluminum, beryllium (Be), as well as PE (polyethylene), if the latter is allowed from the point of view of fire prevention. Therefore, the presumed moderator/target/reflector(mirror) system is located between the reactor shell and the actual radiation shield. It is preferred if the components of this three-component system are arranged in their own separate cassettes because in this way any of the components can be mobilized independently of one another; this is, however, not a requisite. Furthermore, said three-component cassette system can be arranged within a common container provided with extra radiation shielding. The above technique and embodiment can be made use of in the case of irradiation channels of (research) reactors, as well; here the production batch will be much smaller, however, the product can be prepared in a shorter period of time. As far as budget is concerned, in deployment next to the reactor shell is applied, said “waste” neutrons will do the job free of charge, contrary to the case of the channel-type embodiment which is rather recommended by way of experiment, as well as for smaller production amounts and/or for research purposes. It is a requirement, however, that any of the mother elements (in any combination and/or composition) specified in the object of the invention are contained by the target in the amount of at least 8.0 weight %. Some possible examples when W is selected are as follows: W90/Ti10, W75/W25, W90/Cr10, W60/Cu40, W90/Ag10, W75/Re25, etc., as well as WC, WO3, WSi2, but—as it was mentioned earlier—W2B is not recommended, while WS2 is not preferred. It is noted that if the material of the target also contains moderator and/or reflector(mirror) components, said components should not actually be taken into account as target. The product will be basically a specific alloy, i.e. the mixture of the mother element and the daughter element(s), since these species can be actually alloyed with one another. In the same process, it is also possible to activate the daughter and mother elements further so as to produce secondary daughter elements, such as e.g. by the process of W→Re→Os, as will be discussed later in more detail. Reverting now to a detailed description of the cassettes, in the case of the irradiation channel construction, the important factor is apparently the inner diameter of the channel which is, in general, about 10 cm in size. In the case of the reactor shell, an embodiment of the cassette type with a base plate of e.g. 90 by 90 cm in size is preferred, however, this represents only a possible example. The base material of said cassettes can be Mg, Al, Fe, Zr, as well as any suitable alloys thereof. The cassettes can be grouped in three, such as moderator/target/reflector(mirror), wherein each group (cassette) is separated from the others. It is highly preferred if the respective cassette of the target can be taken out separately from amongst the two others. Separate displaceability of the target cassette is also preferred, as the moderator cassette and/or the reflector cassette have to be displaced much rarely. Apparently, the displacement of said cassettes is performed by robots and manipulators. In what follows, the cassette types and the container are discussed in more detail. (a) The moderator cassette is mostly determined by the neutron spectrum and flux of the reactor. It is an object to provide a thermal reactor neutron yield that is maximal at the exit side of said cassette. It is noted that most reactor types produce enough thermal neutrons to activate the target without even a moderator, however, this is a slower process. The moderator can be provided by reactor grade C graphite, H2O, D2O, paraffin and He. When paraffin is used, to moderate fast neutrons and (reactor) neutrons a thickness of about 40 cm and about 20 cm, respectively, thereof is required. For C graphite, the thickness should be about 10 cm (this is considered to be the most advantageous). (b) The target cassette is filled up with one of the (perhaps more) mother elements mentioned before. When selecting the material thickness, self-absorption of the target element(s) and that of the resultant daughter element(s) have to be taken into account. Thus, the recommended material thickness ranges preferably from 10 to 15 cm. It is preferred to form the cassette with a net volume of 100 dm3. Depending on its filling, the cassette has a gross mass of 2 to 4 tons. (c) The reflector(mirror) cassette is constructed with similar principles in mind; however, the backscattering of neutrons has to be considered with a thermal value. The usage of Be is preferred, but due to its intoxicating nature, rather BeO is recommended. Due to its hydrogen content, PE is a scatter medium, however, it is not heat-resistant. Mainly Ni and Fe, as well as any suitable alloys thereof, and/or Bi, Pb (not preferred too much), Bi2O3 which is stable, heat-resistant and chemical resistant enough can be offered, too. (d) The (three-component) cassette supporter container, as is also reflected by its name, is a means suitable for holding the three cassettes together. As far as its base material is concerned, it is identical to those of the cassettes. Moreover, it is mechanically designed so as to withstand to chemical, thermal and mechanical damages and also to be less activable as structural material. It is also equipped with suitable means and elements for effecting displacements and connections. Its dimensions are preferentially about 90 by 90 by 60 cm; this corresponds well to the sizes expressed in units of inches well-spread and used in the international practice. Except its side facing to the reactor (i.e. the front side), said container can be provided with extra radiation shield. The gross mass of the container with the cassettes is about 8 to 10 tons. It is noted that in the case of the irradiation channel the situation is simpler: the thermal neutron flux can be affected ab ovo by means of the built-in filtering means of the reactor. Hence, it is not sure that there is a need for the moderator cassette, which is apparently a cylindrical casing, in this case. The construction in principle follows that of the system with cassettes, however, as here there are provided means of much smaller weight, the base material of the casing can be Al and/or Fe. The length of said cylindrical casing corresponds to the width of said cassettes. This means that preferably and purposively each casing is 10 to 20 cm in length. Since in this case there is no need for the container support, a radial shield cassette can also be arranged after said reflector(mirror) cassette as a fourth component. In what follows, the present invention is overviewed with reference to the FIGURE. I. (Reactor)neutrons 11 leaving through the reactor shell 1 passes over the front side of container 2 and then enter the moderator cassette 3 containing suitably chosen moderator substance 4. From here they proceed with a maximal thermal neutron yield 13 and enter the target cassette 5, and the target 6 mother element. The remaining thermal neutrons 12 pass further and enter the reflector(mirror) cassette 7 containing suitably chosen reflector(mirror) substance 8 that scatters part of the thermal neutrons 12 entering here back towards the target 6. The container 2 itself, except its front portion, is equipped with extra radiation shield 9 which is protected by an outer envelope 10 that is preferably based on Fe. II. The irradiation channel requires no further explanation. Reverting now to the prior art techniques and technology, the excellent work of C. Rubbia (PCT/EP97/03218, filed on 19 Jun. 1997.) should be here also mentioned, which exploits neutrons escaping from a reactor, but makes use of other neutron sources as well. This is preferred mainly when existing radioactive (power plant) wastes are to be activated further so as to transmute them into elements of lower half-lives. The author also discloses—amongst others—the producibleness of various (medical) isotopes, the doping of Si and Ge based elements with impurities, etc. It is essential, however, that the transformation (transmutation) of lanthanides and platinum metals is not mentioned amongst the objectives of the invention. Although the author has constructed a table collecting all the elements and their isotopes from Na to Th which could be produced by the apparatus of the author, said apparatus is not descriptive—and, hence, is not meritorious—to the system comprising cassettes and a container in accordance with the subject-matter of present invention as disclosed here. Reverting now, with reference to some highlighted examples, to the major radiation physics features of transmutation (element transformation) according to the present invention, said examples are numbered in harmony with the tables, wherein the signals “a”, “c” and “e” always refer to mother elements, while the signals “b”, “d” and “f” refer to daughter elements, except the case of Re that can be both a mother and a daughter element (see later), i.e. the transformation process of mother element→daughter element is referred to e.g. by the notation of “a→b”. The atomic number in front of the chemical symbol of a given element, possible isotopes of the element (below said symbol) and the natural abundance ratios thereof within said element in % units, the thermal neutron capture cross-section of each isotope in barn units (rounded values), the half-life (T½) of each isotope, and the types of radiation characteristic of the isotopes are also given (α, e±, γ, K; here K stands for the characteristic radiation, wherein various types of electron irradiations are denoted in a unique form. The state also determines the way of decay, i.e. the mother element transforms into an other element having its atomic number decreased by one). The so-called nuclear isomers are also denoted by the label “m”. wherein Promethium has got no stable isotopes TABLE 1a%barnhalf-life (T½)radiation60Nd4814227.111814312.1724014423.8551*1015yearsα1458.306014617.22214711.0dayse− γ1485.7341491.8hourse− γ1505.62215112.0minutese− γNote:natural Nd also contains an α-radiator; similar elements are Sm, Gd, Hf, Pt, Pb, Th and U. During the transmutation, Nd144 becomes remarkably enriched (as Nd143 isotope has got high neutron-capture cross-section) and Pm isotopes will form. TABLE 1b%barnhalf-life (T½)radiation61Pm601472.6yearse−14953.0hourse− γ1511.1dayse− γ The transmutation reactions, in principle, are the following: a, Nd147→Pm147→Sm147→Eu147. b, Nd149→Pm149→Sm149. c, Nd151→→Pm151→Sm151→Eu151. From this, in practice Pm147 can be utilized, which is pure e−-radiator (0.225 MeV) and will “stabilize” as 62Sm147 which is pure α-radiator with the half-life of 1.2*1011 years (2.23 MeV). Here, the product can be enriched in Nd147/Pm147 isotopes to an extent of about 10% to 15%. TABLE 2a%barnhalf-life (T½)radiation62Sm58201443.092145340.0daysγ K1465*107yearsα14714.97871*1010yearsα14811.2414913.83408101507.441511400093.0yearse− γ15226.7214015347.0hourse− γ15422.71515523.5minutese− γ156 Due to its very high neutron-capture cross-section, Sm151 will be activated further, and thus the formation of Eu151 is not characteristic; it is thought that Eu153 will become enriched within the Sm153 target and/or the transmutation of Eu155-64Gd155 can be detected from Sm155 in traces. TABLE 2b%barnhalf-life (T½)radiation63Eu440015147.821700 152m9.2hourse± K152620012.2yearse± γ K15352.18440154169016.0yearse− γ155158001.7yearse− γ The isomer state of Eu152m will finally stabilize as 64Gd152. Altogether, the Sm153→Eu153 product state can be selected along with an Eu concentration of about 20% to 25%. The transmutation reactions, in principle, are the following: a, Sm145→Pm145→Nd145. (As Sm145 undergoes K-decay.) b, Sm151→Eu151. c, Sm153→Eu153. d, Sm155→Eu155→Gd155. TABLE 3a%barnhalf-life (T½)radiation68Er1601620.13216375minutesγ K1641.56216510hoursγ K16633.41 167m2.5secondsγ16722.9016827.0721699.5dayse− γ17014.8891717.8hourse− γ Due to the K-radiation of Er, only Ho can form in an Erbium target up to Er165. The range of Er166 to Er168 is favorable for us; here the Er168 isotope will become remarkably enriched that slightly compensates for the low cross-section (in barns). TABLE 3b%barnhalf-life (T½)radiation1679.6daysγ K16887.0dayse− γ K69Tm130169100.00130170170129.0dayse− γ K1711.9yearse− γ Altogether, in the transmutation process of Er169→Tm169 even 50% of Er can transform into the state of Tm 169. Ho and Yb will appear in the alloy in a few %. The transmutation reactions, in principle, are the following: a, Er163→Ho163→Dy163. b, Er165→Ho165→Ho165. c, Er169→Tm169. d, Er171→Tm171→Yb171. It should be here noted that this process has already been discussed in the patent document cited previously, and hence the following serves merely as a reminder. TABLE 4a%barnhalf-life (T½)radiation70Yb371680.1312400 169m46.0secondsγ16931.8daysγ K1703.0317114.3117221.8217316.1317431.8460 175m0.0secondsγ175101.0hourse− γ17612.73 177m6.5secondsγ1771.9hourse− γ The stabilizing process of Yb169m→169 leads to Tm169; this process is a direct consequence of the high cross-section value (in barns) of Yb168 and K-decay of Yb169. Lu can form if the process of Yb175m→175 takes place; the formation of other Yb isotopes is not probable. TABLE 4b%barnhalf-life (T½)radiation 174m90.0daysγ174163.0daysγ K71Lu10817597.4035 176m37hourse− γ1762.6040002*1010yearse− γ1776.7dayse− γ In the alloy of the product, Lu can become enriched up to at least 50%; the impurities can be Tm and Hf. The theoretical transmutation reactions are the following: a, Yb169→Tm169. b, Yb175→Lu175. c, Yb177→Lu 177→Hf177. Note: besides the above exemplified reaction processes, it is also possible to produce other lanthanides as well, see e.g. the already mentioned Gd→Tb element transmutation. Hence, as it is already known: TABLE 5a%barnhalf-life (T½)radiation75W181800.1310181145daysγ K18226.4120 183m5.3secondsγ18314.401118430.642 185m1.6minutesγ18573.2dayse− γ18626.4134187901.0dayse− γ W184 becomes enriched in the activation process, however, the transmutation process of W185→Re185 undergoes with low efficiency; on the contrary, the process of W187→Re187 is much favorable. Due to the K-decay of W181, Ta contamination forms; moreover, as a consequence of Os188m→Os188, the rhenium daughter element will contain Os188. TABLE 5b%barnhalf-life (T½)radiation75 Re8418537.07120 186m1.0hoursγ18688.9hourse− γ K18762.93696*1010yearse− 188m18.7minutesγ188218.0hourse− γ It should be here noted that the e−-emission of Re187 is very low both in intensity and in energy. The theoretical transmutation reactions are the following: a, W181→Ta181. b, W185→Re185. c, W187→Re187. If the object is to produce Os, tungsten can be activated further:W→Re→Os (in harmony with the interpretation of 5a→5b→6b) This process is extremely advantageous and economical in the case of the reactor shell type technologies. As it was already mentioned, the Ta181 component will appear in the product in a minimal amount, the major part of rhenium will be Re187 isotope, while the osmium is typically formed by Os188. (This latter can form as much as 10% to 20% of the product.) Osmium can be produced from natural rhenium itself in a more efficient way: TABLE 6b%barnhalf-life (T½)radiation76Os151840.0220018593.6daysγ K1861.591871.64 188m26.0daysγ18813.30 189m5.7hoursγ18916.10 190m10.0minutesγ19026.4040 191m14.0hoursγ191816.0dayse− γ19240.95219360030.6hourse− γ1941.9yearse− The activation of Re185 into Re186m-186 will stabilize by e−- and K-decays as W186 and Os186 isotopes in such a way that the Os portion will be higher. (That is, the initial amount of 1.59% of Os186 increases.) The theoretical transmutation reactions are the following: a, Re186→Os186+W186. b, Re188→Os188. There is no Os185 within the product; other parts of the spectrum are of extremely low intensity. Within the product, Ir can also be present in traces. It is mentioned here that the most valuable stable isotope of natural Os is Os 187 that forms 1.64% of natural Os. The osmium product obtained by the inventive process is a mixture of isotopes Os186 and Os188 and isotopes Re185 and Re187. In what follows two different ways are offered to produce the isotope Os187 from this: (a) in the (n, 2n) reaction of reactor neutrons, the cross-section of Os186 is 0.04 barn, while that of Os188 is 0.005 barn. (Hence, Os188 can transform into Os187, while Os186 remains also a stable isotope. A portion of Re stabilizes as W, a further portion thereof stabilizes as Os186.) (b) by means of intermediary resonance neutrons with energies ranging from 1 eV to 100 keV, the processes of Os186→Os187 and Os188→Os189m→Os189 can be induced; here the state of Re barely changes. The Os isotopes forming here can be separated only by means of complicated separation techniques in any variants. Amongst the platinum metals, producibility of rhodium is going to be discussed in more detail; it is, however, apparent to a person skilled in the art that, besides the elements disclosed previously, it is possible to produce other elements as well. TABLE 7a%barnhalf-life (T½)radiation44Ru3965.531972.9daysγ K981.879912.7210012.6210117.0710231.61110339.7dayse− γ10418.58110514.51hourse− γ TABLE 7b%barnhalf-life (T½)radiation45Rh150 103m57.0minutesγ103100.00149 104m9004.4minutesγ1044542.0secondse− γ 105m45.0secondsγ10535.0hourse− γ 106m2.2hourse− γ10630.0secondse− γ The activation of ruthenium takes place with quite low efficiency. Via K-decaying, Ru97 goes into the state of Tc97m→Tc97, which is a K-radiator isotope with a long half-life (2.6*106 years). The theoretical transmutation reactions are the following:Ru102(n,γ)→Ru103→Rh103m→Rh103. Rh can be easily activated, and thus Tc and Pd contaminants/alloying elements form in the product besides the Ru—Rh alloy. Considering the fact that the products are radioactive, in what follows the energy of the gamma spectrum (MeV) and the specific irradiation power kγ (in relative values) are also given for those isotopes, wherein the number of γ quanta exceeds 10 per 100 decays. Here, the mother element→daughter element transformation reactions are referred to by the label of the type “c→d”. For the various elements, the radiation characteristics and parameters are the following: TABLE 1cTABLE 1dNdMeVkγ→PmMeVkγ1440.090.81510.060.70.530.101510.113.30.120.140.25 TABLE 2cTABLE 2dSmMeVkγ→EuMeVkγ145 0.060.0152m0.121.30.85153 0.100.21520.125.0155 0.100.41540.126.20.730.871.001.011.28155 0.060.80.080.100.12 TABLE 3cTABLE 3dErMeVkγ→TmMeVkγ167m0.210.51710.111.80.300.31 There is no remarkable γ radiation from Tm in the product. TABLE 4cTABLE 4dYbMeVkγ→LuMeVkγ1690.06 1.21760.20 2.70.110.300.130.180.200.311770.12 0.4 TABLE 5cTABLE 5dWMeVkγ→ReMeVkγ183m 0.10 0.51860.140.10.110.161880.150.41870.07 2.80.480.68 TABLE 5a, 5b!TABLE 6dReMeVkγ→OsMeVkγ1860.140.11850.644.11880.15 0.4 TABLE 7cTABLE 7dRuMeVkγ→RhMeVkγ1030.501.2104m0.051.01050.673.9105m0.130.10.72106m0.2213.40.410.450.510.620.720.740.821.051.141.221.54 A possible and well-known technique to separate the mother and daughter elements of the product is to keep the product in a molten phase by means of maintaining it at the requisite temperature until the element components get separated from one another, driven by gravity, due to the difference in their densities. If the product is a powder, it can be oxidized; in particular lanthanides are stable in the forms of LaF3 and La2O3, wherein the latter oxidized form is recommended. (Preferably, the crucible is provided by a vertical ceramic tube, the inner surface of which is coated with any of AL2O3, Ta, W and Ir according to needs. For oxide melts, the most preferred is Ir, while for metallic melts Ta and W are recommended.) Processing of the products seems to be the simplest in the oxidized form. (Note: B.P.=Boiling Point, M.P=Melting Point, D.=Density) Here, the mother element→daughter element transformation reactions are referred to by the label of the type “e→f”. Nd→ Pm;Sm→ EuB.P.: 3068?17911597M.P.: 102110271077822D.: 7.00?7.545.24Nd2O3→ Pm2O3Sm2O3→ Eu2O3M.P.: 2272?23502056D.: 7.24?7.438.18Table 1eTable 1fTable 2eTable 2f Er→ Tm;Yb→ LuB.P.: 2863194711943395M.P.: 152915458191663D.: 9.059.326.989.84Er2O3→ Tm2O3Yb2O3→ Lu2O3M.P.: 2400instable2346instableD.: 8.648.909.179.41Table 3eTable 3fTable 4eTable 4f W→ Re;Re→ OsB.P.: 5660562756275027M.P.: 3410318031803045D.: 19.321.021.022.5W O3Re2O7Re2O7Os2O3M.P.: 1473297297instableD.: 7.168.208.20?Table 5eTable 5f5f ! Table 6eTable 6f Thus, the Os isotopes of the product will basically consist of merely Os186 and Os188 isotopes. To separate the mother and daughter elements of the product, the same technique is recommended as in the case of the lanthanides: Ru→RhB.P.: 39003727M.P.: 23101965D.: 12.2012.40Table 7eTable 7f Oxidized forms thereof (in practice) are not known. It should be here noted that the neutron-capture cross-section of the daughter element Rh forming in the process Ru→Rh is much higher than that of the mother element Ru. Consequently, it decays further upon activation, wherein the half-lives of said decays are relatively short. Taken the decay and forming factors of the mother and daughter element(s), as well as the activation time and the half-lives also into account, there will be no daughter element Rh present in the product if the concentration of Ru within the target does not exceed the value of at least 8 weight %, since the target cannot be transmuted into Rh even if it contains 100% Ru. Namely, upon reaching a so-called radioactive decay balance, the activity of the daughter element is at maximum and hence further activation of the mother element is no longer preferred which means that there will always be mother elements that have undergone no transmutation. (This statement is even more relevant when Os187 is produced.) Briefly summarized: the teaching related to the production process of the products discussed here is considered to be a proof of applicability of the present invention in practice. The technology based on the utilization of a “cassette—container plus reactor shell” type arrangement illustrated above enables a significant industrial increase. This increase will induce further changes in those industrial segments as well where the inventive solution becomes applied thereby affecting/changing the future of these segments, too. Jász Árpád-Lengyel Tamás: Izotóplaboratóriumi zsebkönyv Müiszaki Könyvkiadó 1966. Neutron Cross section—Brookheaven National Laboratory, 2nd edition 1958. S. F Mughabghab et al. Neutron Cross section: Neutron Resonance Parameters and Thermal Cross Section v.1 (Neutron Cross sections Series) (Vol1) Saunders College Publishing Nuclear Fission and Neutron-included Fission Cross-section (Neutron physics and nuclear data in science and technology), Pergamon Press 1981 Atlas of Neutron Resonances, 5th edition: Resonance Parameters and Thermal Cross Sections. Z=1-100 S. F. Mughabghab, Elsevier Science (5th edition 2006) |
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062563637 | description | SPECIFIC DESCRIPTION As seen in FIGS. 1 and 2 a container 1 for spent nuclear fuel elements 2 (FIG. 5) has a massive and cylindrical side or outer wall 3 centered on a normally upright axis 21 and defining an interior 4. A floor plate 5 closes the bottom of the wall 3 and a cover 6 the top. The walls 4, floor 5, and cover 6 can include shielding materials, for instance steel with polyethylene shielding. A rack or basket 7 forming a plurality (here 32) of axially full length wells or passages 8 and 8a is itself formed as a stack of levels 15 each formed in turn by neutron-absorbing borated aluminum plates 9 and 9a joined together at slide joints 10 (FIG. 4). Snugly fitted in each of the wells or passages 8 and 8a is a square-section stainless-steel tube 11 holding the elements 2 and extending the full axial length of the interior 4 of the container 1. The plates 9a each extend diametrally between outer ends 14 that bear against an inner surface of the side wall 3 so as to be in good heat-transmitting contact therewith. These plates 9a, which are some 12 mm thick, subdivide the interior 4 at each level 15 into four quadrants 22, 23, 24, and 25. The plates 9 are some 8 mm thick and have lengths equal to two or three times the side length of the identical square-section wells 8 and 8a. They fit together at joints 16 where they are formed with half-width slots 17 that fit together to form the inner wells 8 that are wholly defined by the plate 9 and 9a and identical outer wells 8a of which at most two or three sides are formed by the plates 9 and 9a. Around the rack 7 the container 1 is provided with filler blocks 12 each made of cast aluminum in which are imbedded lead shield elements 13. Each block 12 has a part-cylindrical outer surface and an inner surface formed as one or two planar sides, so that together with the rack 7 they form the outer walls of the outer wells 8a. The rack 7 is assembled right in the side wall 3 one level 15 at a time, each level 15 being formed by two plates 9a, eight short plates 9, eight long plates 9a, and twelve filler blocks 12. Once complete, the square-section element-filled tubes 11 are slipped into place. As shown in FIG. 5 each such tube 11 carries a multiplicity of spent fuel rods 18 and is formed with an array of guide passages or tubes 19, here lying in a uniform array, each holding a respective absorber rod 20 to maintain the charge subcritical. |
summary | ||
description | The present application claims priority under 35 U.S.C. §119(e)(1) to U.S. Ser. No. 61/541,154, filed on Sep. 30, 2011. The present application also claims priority under 35 U.S.C. §119 to European patent application serial number 11 007 980.3, filed on Sep. 30, 2011. These contents of these applications are hereby incorporated by reference in their entirety. The disclosure relates to charged particle beam systems and methods of operating charged particle beam systems. The disclosure particularly relates to performing line scans using a charged particle system. Conventional charged particle beam systems typically include a beam source and a focusing lens to direct the charged particle beam onto an object, a beam blanker to blank and un-blank the beam and a beam deflector to scan the beam across the object. Charged particle beam systems may further include a detector to detect charged particles and/or radiation emerging from the object due to the incident beam. Using the detector, images of the object can be generated by scanning the beam across the object and associating detected particle intensities with corresponding scan locations. Scanning of the beam across the object typically includes performing a plurality of line scans, wherein the location of incidence of the beam on the object is continuously moved along straight paths. A line scan is typically initiated by a trigger signal. Other desired actions, such as starting the scan deflection, ending the scan deflection, un-blanking the beam, blanking the beam, starting data acquisition and stopping data acquisition, are electronically controlled relative to the trigger signal by providing adjustable delay circuits in the respective electronic circuits controlling the beam deflector, the beam blanker and the data acquisition, respectively. Other charged particle systems can be used, for example for writing patterns on the object by deflecting the beam to a location within a pattern feature to be written, un-blanking the beam and further deflecting the beam such that it is incident on other locations within the pattern feature. The disclosure is based, in part, at least, on the realization that adjusting multiple electronic delay circuits for performing various operations can be tedious, lacking in reproducibility and inflexible. According to embodiments of the present disclosure, a method of operating a particle beam system includes digitally controlling a first digitally controlled module of the particle beam system and a second digitally controlled module of the particle beam system, sending digital command data to the first and second digitally controlled modules, wherein the digital command data includes at least a first command for the first digitally controlled module and a second command for the second digitally controlled module, wherein the digital command data is generated based on information representing a time when the first command is to be executed by the first digitally controlled module and on information representing a time when the second command is to be executed by the second digitally controlled module. The first and second digitally controlled modules can, in particular, be any of a beam deflector and a beam blanker, a first and second beam deflectors, a beam deflector and a signal detector, and a beam blanker and a signal detector. According to some embodiments, a method of operating a charged particle beam system includes determining at least one deflection amount and at least one deflection time, generating a first digital command representing an instruction for a beam deflection module of the particle beam system connected to a data network of the particle beam system to provide the at least one deflection to a charged particle beam corresponding to the deflection amount, and sending the digital command to the data network such that the beam deflection module can receive the digital command data in order to perform the instructed operations, i.e. to provide the deflection corresponding to the deflection amount at the deflection time. The method may further include determining a beam un-blank time and a beam blank time, generating second digital command data instructing a beam blanking module of the charged particle system and connected to the data network to un-blank the charged particle beam at the beam un-blank time and to blank the charged particle beam at the beam blank time and sending the second digital command data to the data network such that the beam blanking module can receive the digital command data in order to perform the instructed operations. According to particular embodiments, the method may further include generating a data structure including plural data records, wherein each data record includes a command representing an instruction for at least one of the beam deflection module and the beam blanking module, and a command time representing a time at which the instruction is to be sent to the data network; sorting the records of the data structure by command time; and sending a set of digital commands encoding the commands included in the data records to the network in an order corresponding to an order of the sorted records. According to some embodiments, the method includes determining first and second deflection amounts and first and second deflection times. For example the first and second deflection amounts and times may be used to instruct a line scan, starting at the first deflection time with the first deflection amount and ending at the second deflection time with the second deflection amount, wherein the deflection is changed continuously or in discrete steps during the time period between the first deflection time and the second deflection time. According to particular embodiments herein, at least one digital command of the set represents a combined instruction for the beam deflection module to provide a deflection to the beam corresponding to the first deflection amount and to subsequently provide a deflection to the beam corresponding to the second deflection amount. The at least one digital command may include, for example, at least one data element representing a time difference between the second deflection time and the first deflection time. Also, the at least one digital command may include at least one data element representing at least one of a deflection step size by which the first beam deflection module is to change the deflection of the beam in subsequent time steps, and a number of steps in which the first beam deflection module is to change the deflection of the beam between the first deflection time and the second deflection time. According to further embodiments, the method further includes determining a data acquisition start time and a data acquisition stop time, generating digital commands instructing a data acquisition module of the charged particle system and connected to the data network to start collecting data representing detected particle intensities at the acquisition start time and to stop collecting digital signals representing the detected particle intensities at the acquisition stop time and sending these digital commands to the data network such that the data acquisition module can receive the digital command data in order to perform the instructed operations. One or more data records of the sorted data structure may then include a command representing a corresponding instruction for the data acquisition module. The data network is, within the present disclosure, a communication device supporting transfer of digital data between modules connected to the communication device. The network can be configured to have a particular topology, such as, for example, point-to-point, bus, star and ring. According to certain embodiments, the set of digital command data is generated such that at least one digital command of the set represents both an instruction for the beam deflection module to provide the deflection to the beam corresponding to the deflection amount and an instruction for the beam blanking module to un-blank the beam. In exemplary embodiments herein, plural digital commands instructing different modules to change their state, such as the beam deflection module to change the provided deflection or the beam blanking module to change from blanking the beam to un-blanking the beam, may also contain instructions for other modules to maintain their state. These latter instructions have no effect on the other modules, but allow for a uniform format of the digital commands and easy distribution to the modules connected to the network in a broadcast type protocol. Individual digital command data sent across the network can be sent as one packet, or they can be split to fit into plural packages. Irrespective of whether individual command data are split into plural packages or not, they can be represented as a buffer or a set of plural bits representing one or more data elements. In an individual digital command data buffer, at least one data element identifies a command to be performed by the addressed module connected to the network. For example, the data element may represent the command “un-blank the beam” or “blank the beam” for execution by the beam blanker, or “start scan” or “stop scan” for execution by the beam deflector. The digital command data may further include one or more data elements representing command parameters providing additional information for execution of a particular command. For example, the command “start scan” may be accompanied by one or more parameters representing a duration of the scan or a number of scanning steps and a time duration for which the beam should remain at a same scan position during the scan. According to embodiments, the present disclosure provides a particle beam system including at least a first digitally controlled module, a second digitally controlled module, and an encoding module configured generate digital command data, wherein the digital command data includes at least a first command for the first digitally controlled module and a second command for the second digitally controlled module, wherein the digital command data is generated based on information representing a time when the first command is to be executed by the first digitally controlled module and on information representing a time when the second command is to be executed by the second digitally controlled module. The first and second digitally controlled modules can, in particular, be any of a beam deflector and a beam blanker, a first and second beam deflectors, a beam deflector and a signal detector, and a beam blanker and a signal detector. According to an exemplary embodiment, a particle beam system includes a charged particle beam source configured to generate a charged particle beam; a data network; a beam blanking module connected to the data network and configured to blank and un-blank the charged particle beam; a focusing lens configured to focus the charged particle beam onto an object; a beam deflection module connected to the data network and configured to deflect the beam; a calculation module configured to determine a deflection time, a beam un-blank time and a beam blank time, to generate a data structure including plural data records, wherein each data record includes a command representing an instruction for one of the beam deflection module and the beam blanking module, and a command time representing a time at which the instruction is to be sent to the data network, and to sort the data records of the data structure by command time; and an encoding module configured to generate a set of digital commands encoding the commands included in the data records and sending the generated digital commands to the network in an order corresponding to an order of the sorted records. In the exemplary embodiments described below, components that are alike in function and structure are generally designated by alike reference numerals. Therefore, to understand the features of the individual components of a specific embodiment, the descriptions of other embodiments and the summary may be referred to. FIG. 1 is a schematic illustration of an embodiment of a particle beam system, which is an electron microscope in the illustrated example. The particle beam system 1 includes a particle beam source 3 including a cathode 5, a suppressor electrode 7 and an extractor electrode 9 for generating an electron beam 11. The electron beam 11 emerging from an opening in the extractor electrode 9 is accelerated by an anode 13 to a predetermined kinetic energy and enters a beam tube 15 via an opening in the anode 13. The electron beam may traverse a condenser lens 17, an aperture 19 provided in an electron detector 21. The electron beam further traverses an objective lens 23 for focusing the electron beam 11 at a location 25 in an object plane 27 of the objective lens 23. A surface of an object 29 which is to be inspected or manipulated with the particle beam system 1 can be arranged in the object plane 27. The objective lens 23 includes an annular coil 31, which is arranged in an annular pole piece 33, which includes an annular upper pole piece 35 and an annular lower pole piece 37 such that a magnetic field focusing the electron beam 11 is generated in an annular gap between the two pole pieces 35 and 37. The objective lens further includes a terminal electrode 39 which is arranged spaced apart from a lower end 41 of the beam tube 15 and has an opening traversed by the electron beam 11. An electric field generated between the lower end of the beam tube 41 and the terminal electrode 39 decelerates the electrons, propagating inside the beam tube 15 at a high kinetic energy, to a desired lower kinetic energy at which they are incident on the object 29. This electric field may provide an additional focusing effect together with the magnetic field. The individual components of the particle beam system 1 are controlled by a controller 42. The controller is shown in FIG. 1 as a functional block and includes plural control modules which can be spatially separated from each other or arranged together in, for example, a housing. Also, one or more of the plural control modules can be embodied as individual electric circuits, and/or they can be embodied as software modules running on a suitable processor, such as a general purpose processor, together with other control modules or modules performing other tasks, such as providing a user interface to the system 1. One module of the controller 42 controls the beam source via connectors 43 for supplying a heating current to the cathode 5 and defining a potential of the cathode. Electric potentials of the suppressor electrode 7 and the extractor electrode 9 are controlled via connectors 44. An electric potential of the beam tube 15 and the anode 13 is defined by the controller via a connector 45. For this purpose, the controller 42 includes a stabilized high voltage source, which supplies a voltage of, for example, 8 kV with respect to ground to the connector 45. Beam deflectors 47 which are controlled by a beam deflection module of the controller 42 via connectors 48 can be arranged in the objective lens 23. The beam deflectors can be magnetic beam deflectors which may provide adjustable deflecting magnetic fields within the beam tube 15 in order to vary the location 25 at which the electron beam 11 is incident on the object 29, and to scan the particle beam 11 across a portion of the surface 27 of the object 29. The particle beam 11 incident on the object 29 causes secondary electrons or backscattered electrons to emanate from the object 29. A portion of these electrons may enter the beam tube 15 and can be detected by the electron detector 21. An exemplary trajectory of a secondary electron incident on the electron detector 21 is labeled with reference numeral 51 in FIG. 1. Detection signals triggered by incident electrons are output by the electron detector 21 at a connector 53 and can be read in by a data acquisition module of the controller 42. The particle beam source 3 is preferably operated in a stationary mode, i.e. once it is put into operation, the particle beam source 3 is operated for several hours or even days under constant conditions such that the electron beam 11 is continuously generated. However, it is desirable to not allow the electron beam 11 to be constantly incident on the object 29 and to be able to switch the beam on and off as desired. For this purpose the particle beam system 1 includes a beam blanker system 55 which includes a pair of deflector electrodes 56, 57 which can be arranged inside the beam tube 15 such that the electron beam 11 traverses a gap formed between the deflector electrodes 56, 57. The controller 42 includes beam blanking module which supplies electric potentials to the deflector electrodes 56, 57 via connectors 58 and 59. If both deflector electrodes 56, 57 are at the same electric potential, the beam 11 traverses the gap between the deflector electrodes along a straight line. Preferably, the deflector electrodes are at a same electric potential as the beam tube 15. If the deflector electrodes 56, 57 are at different electric potentials, an electrostatic field is produced between the two deflector electrodes. This electric field deflects the electron beam 11 away from its original trajectory. The deflected electron beam is shown in FIG. 1 as a broken line 11′ and is incident on a plate 61 arranged in the beam tube 15. The plate 61 has an aperture 62 which is traversed by the non-deflected beam 11 to be incident on the surface 27 of the object 29. The electron beam 11′ incident on the plate 61 is absorbed and cannot reach the surface 27 of the object 29. FIG. 2 shows charts illustrating a time sequence of actions performed by the beam deflection module, the beam blanking module and the data acquisition module of the controller 42. These actions are performed within a procedure to record an image using the electron microscope shown in FIG. 1. Recording an image involves recording plural lines of image information by scanning the beam along lines and recording corresponding detected particle intensities. The particular actions illustrated in FIG. 2 are related to performing one such line scan. Chart (a) represents a current I supplied to the beam deflectors in dependence of time. This current is at a constant first level 101 in order to provide a deflection corresponding to a first deflection amount in the beginning. The beam scanning starts at a time Tss and stops at a time Tse, such that the current is at a different constant second level 105 in order to provide a deflection corresponding to a second deflection amount after time Tse. Between times Tss and Tse, a continuous increase of the current level occurs, as indicated by reference numeral 106, in order to continuously change the provided deflection such that a line scan is performed in the time period between times Tss and Tse. The currents according to lines 101, 106 and 105 are generated by the beam deflection modules upon receipt of corresponding commands. A command to start the scanning is sent to the network and received by the beam deflection module at a time Tcss before time Tss. A time difference δT which is the difference between time Tss and Tcss corresponds to an internal processing time of the beam deflection module. This processing time δT is predetermined and known, such that the time Tcss can be suitably selected such that the physical beginning of the beam deflection occurs at time Tss. Similarly, a command to stop the scanning is sent to the network and received by the beam deflection module at a time Tcse. Again, time Tcse is earlier than time Tse, wherein a time difference δT between time Tse and time Tcse, accounts for a processing time for the beam deflection module to stop scanning. This processing time δT subsequent to Tcse may have a same duration or a different duration than the processing time δT subsequent to Tcss. Chart (b) illustrates a time dependency of a voltage applied to the electrodes 56, 57 of the beam blanker. In this example, the un-blanking of the beam occurs at a time Tub, and the blanking of the beam occurs at a time Tb. Since the beam blank module needs some time to execute received commands and to change the voltages applied to the electrodes, corresponding commands are sent to the network and received by the beam blanking module at earlier command times Tcub and Tcb, respectively. Herein, a time difference δT1 between Tub and Tcub can be longer than the time difference δT2 between Tb and Tcb. Blanking the beam involves deflecting the beam traversing the aperture 62 by a small amount such that it is incident on the plate 61. This can be quite fast since the beam is un-blanked even before the deflection of the beam provided by the deflector 56,57 has settled to a stable value. On the other hand, un-blanking the beam involves directing the beam, which is initially incident on the plate 61, such that it exactly traverses the aperture 62 after the deflection of the beam provided by the deflector 56,57 has settled to a stable value. This may involve relatively more time. Exemplary values for δT1 and δT2 can be within a range, for example, from 50 ns to 300 ns. In particular, δT1 and δT2 can have different values. For example, δT1 can be longer than δT2. According to some embodiments, the time difference between the blank time and the un-blank time differs from the time difference between the second command time and the first command time by more than 50 ns, more than 100 ns or more than 200 ns. The time Tub at which the beam is un-blanked is, in the illustrated example, earlier than the time Tss at which the beam deflection module starts the scan. This is due to a time for the particles to travel between the electrodes 56, 57 of the beam blanker and the deflection coils 47 of the beam deflector. Due to this traveling time, the beam blanker is operated earlier than the beam deflector. Similarly, the time Tb at which the beam blanker blanks the beam, is earlier than the time Tse at which the beam deflection module stops scanning the beam. Chart (c) illustrates an operation of the data acquisition module of the controller 42. The detector 21 continuously produces analog detection signals irrespective of whether the beam is blanked or un-blanked or scanned. To record an electron microscopic image of an object, detected particle intensities are associated with scanning locations of the beam at the time of recording, i.e. with locations of the object. For this purpose, it is desired to collect a sequence of data values representing detected particle intensities, wherein the sequence starts when the scanning beam is at a corresponding starting position, corresponding to, for example, a left image margin, and the sequence stops when the scanning beam is at a different position corresponding, for example, to a right image margin. The starting and stopping of the data acquisition is synchronized with the scanning of the beam, wherein the data acquisition is delayed relative to the beam deflection due to times for the primary particles to travel from the deflector to the object and the secondary particles to travel from the object to the detector. As shown in chart (c) the data acquisition module starts collecting the digital image data at a time Tas which is later than the time Tss at which the beam deflection module starts scanning. Similarly, the data acquisition module stops data acquisition at a time Tae which is later than time Tse at which the beam deflection module stops scanning. Again, since the data acquisition module involves some processing time for executing commands, a command for instructing the data acquisition module to start collecting data is sent to the network and received by the data acquisition module at a time Tcas which is earlier than Tas, and a command for instructing the data acquisition module to stop collecting data is sent to the network and received by the data acquisition module at a time Tcae which is earlier than Tae. Chart (d) illustrates the time sequence of the commands illustrated above for the present exemplary embodiment: Tcub<Tcss<Tcas<Tcb<Tcse<Tcae. According to other embodiments, other time sequences are possible, depending on, for example, traveling times of particles in the system and processing times of the individual modules. FIG. 3 is a schematic illustration of a portion of the controller 42 used for controlling the beam blanker, beam deflector and data acquisition. For this purpose, the controller 42 includes a calculation module 111 which determines the times Tub, Tss, Tas, Tb, Tse, Tae based on plural parameters. Some parameters are received via an interface 113, which can be an interface to a local area network or a keyboard configured to receive data representing the task to be performed, such as left and right boundaries of an image to be recorded, a pixel speed, an image resolution and other parameters. The calculation performed by the module 111 may take a considerable amount of processing time and takes into account other parameters representing physical properties of the charged particle beam system, such as a kinetic energy and a speed of the charged particles in order to calculate corresponding traveling times, and other parameters. As soon as the module 111 has completed the calculation of the times Tub, Tss, Tas, Tb, Tse, Tae, the corresponding earlier command times Tcub, Tcss, Tcas, Tcb, Tcse, Tcae are calculated based on the processing times used by the respective modules executing the commands. Data representing the Tcub, Tcss, Tcas, Tcb, Tcse, Tcae and additional command parameters are transmitted to a command generation module 115 which encodes the commands and additional parameters into digital command data suitable to be sent to the corresponding modules across a network 117. The module 115 also supplies the generated command data to the network 117 according to the time sequence illustrated in chart (d) of FIG. 2. For this purpose, the module 115 receives a clock signal from a clock 119. The same clock signal is also supplied to components of the network 117 and a beam blanking module 121 to which the electrodes 56, 57 of the beam blanker are connected, a beam deflection module 123, to which the beam deflector 47 is connected, and a data acquisition module 125, to which the detector 21 is connected. The data collected by the data acquisition module 125 are supplied to an image memory 127. The beam blanking module 121, the beam deflection module 123 and the data acquisition module 125 are connected to the network 117 such that they can receive the commands supplied by the module 115. The modules 121, 123, 125 are configured to execute corresponding actions upon receipt of the commands from the network 117. In the present illustration, it is assumed, that the time between sending a command to the network and the reception of the command by the respective module is negligible. However, if this assumption is not sufficiently accurate in practice, a time for the commands to travel across the network can be taken into account when the times Tcub, Tcss, Tcas, Tcb, Tcse, Tcae are calculated based on the times Tub, Tss, Tas, Tb, Tse, Tae. FIG. 4a schematically shows an exemplary layout of data elements within a data buffer encoding an exemplary command. The data buffer includes a number of n bits with low order bits located on the left in FIG. 4a and high order bits located to the right in FIG. 4a. A first number of consecutive bits 141 represents an address within the network of the destination module of the command. Depending on a topology of the network, such address can be omitted. For example, a network having point-to-point topology, would not require that an address of the addressed module is included in the command. A second number of consecutive bits 143 represents the command. In the illustrated example, the command is “begin scanning” instructing the beam deflection module to start scanning. A third number of consecutive bits 145 form a data element representing a command parameter. In the illustrated example, this command parameter is the duration of the scan and instructs the beam deflection module to stop scanning after this duration. As a consequence a separate subsequent command separately instructing the beam deflection module to stop scanning is not necessary. The two commands instructing the beam deflection module to start scanning and instructing the beam deflection module to stop scanning are combined into a single combined command, accordingly. According to other examples, such combined commands are not used, and separate commands are generated starting and stopping the scanning, wherein a data element representing the duration of the scan need not to be included in the command data. A fourth number of consecutive bits 147 form a data element representing a further command parameter which is a number of scanning steps to be performed between start of the scan and end of the scan. Thus, this parameter determines the image resolution to be achieved with the scan. FIG. 4b schematically shows another exemplary layout of data elements within a data buffer encoding plural commands. The data buffer includes a number of m bits with low order bits located on the left in FIG. 4b and high order bits located to the right in FIG. 4b. A first number of consecutive bits 149 represents the command for the beam blanking module to either blank or un-blank the beam. For example, this command can be encoded by one single bit. A second number of consecutive bits 151 represents the command for the data acquisition module to either collect data or to not collect data. Also this command can be encoded by one single bit. A third number of consecutive bits 153 represents the command for the beam deflection module to provide a deflection to the beam corresponding to a given deflection amount. The deflection amount can be encoded, for example, by two sub-groups 155 and 157 of consecutive bits within the third number of consecutive bits 153, wherein sub-group 155 encodes the deflection in an x-direction and sub-group 155 encodes the deflection in an y-direction of the deflection module. The data buffer can be broadcasted simultaneously to all modules, i.e. the beam deflection module, the beam blanking module and the data acquisition module, wherein the beam blanking module extracts bits 149 from the data buffer and process these bits as a received command, the data acquisition module extracts bits 151 from the data buffer and process these bits as a received command, and the beam blanking deflection module extracts bits 153 from the data buffer and process these bits as a received command. If, with such layout, the state of only one module has to be changed by a command, it is sufficient to generate the command for this module based on the desired change, and it is easy to generate the commands for the other modules such that they are identical to previous commands to those modules or to the current states of these modules. If, for example, a first command includes an instruction for a beam blanking module to un-blank the beam and a subsequent command includes an instruction for a beam deflection module to start a line scan while the state of the beam blanking module should remain un-changed, i.e. un-blanked, the subsequent command may contain a repeated instruction to un-blank the beam, since such repeated instruction will not change the current state of the beam blanker. If, according to another example, a first command includes an instruction for a beam deflection module to start a line scan and a subsequent command includes an instruction for a beam blanking module to un-blank the beam while the state of the beam deflection module should remain un-changed, i.e. the beam deflector should continue with the line scan, the subsequent command may contain a new instruction for the beam deflection module to start a line scan, wherein the parameters of the new line scan are selected such that the new line scan steadily continues the previous line scan without interruption. It is apparent that many variations of the command data layout illustrated in FIGS. 4a and 4b are possible. FIG. 5 is a schematic illustration of an embodiment of a particle beam system which is an ion beam system in the illustrated example. The ion beam system 1 includes an ion source 3 and electrodes 7 and 9 for extracting ions from the source 3 and accelerating the extracted ions to generate an ion beam 11 which is collimated by a condenser lens 17. The beam 11 is further accelerated by an electrode 13 and traverses a beam blanker 55 including deflector electrodes 56 and 57, and a plate 61 having an aperture 62 configured such that the beam 11 may traverse the aperture 62 when same electric potentials are applied to the electrodes 56 and 57 and such that the beam 11 is incident on and absorbed by the plate 61 if different electric potentials are applied to the electrodes 56 and 57. An objective lens 23 configured to focus the ion beam 11 in an object plane 27 is provided downstream of the beam blanker 55. The ion beam system 1 further includes a first beam deflector 47 arranged downstream of the beam blanker 55, and a second beam deflector 47′ arranged downstream of the first beam deflector 47 and upstream of the objective lens 23. The beam deflectors 47 and 47′ are configured to direct the ion beam 11 to selected locations within the object plane 27. In the illustration of FIG. 5, the ion beam 11 is focused in the object plane 27 at a distance d of an axis of symmetry 2 of the objective lens 23. To achieve such deflection of the ion beam 11, the first deflector 47 deflects the beam away from the axis of symmetry 2, and the subsequent second deflector 47′ deflects the beam towards the axis of symmetry 2 such that the beam 11 traverses the objective lens 23 close to its axis of symmetry, such that aberrations introduced by the objective lens 23 are maintained at a relatively low level. A secondary particle detector 21 is located close to the object plane 27 such that secondary particles generated by the incident ion beam 11 can be detected. A line 51 in FIG. 5 represents an exemplary trajectory of an electron released from an object and incident on the detector 21. The ion beam system 1 includes a controller for controlling the individual components for generating and directing the ion beam 11 to the object plane 27. Similar to the illustration of FIG. 1, the controller 42 is shown as a functional block including plural control modules which can be physically separated from each other and/or are embodied as software modules running on a suitable processor. In particular, the ion source 3 is connected to the controller 42 via a connector 43 such that the controller 42 can energize and operate the ion source 3. The electrodes 7, 9, 13 and 61 are connected to the controller 42 via connectors 44 and 45 such that the electric potentials applied to the electrodes can be adjusted by the controller 42. Similarly, connectors 49 are provided to connect lenses 17 and 23 to the controller 42 such that the controller 42 can supply suitable electric potentials and currents to the lenses in order to collimate and focus the ion beam 11. The first and second beam deflectors 47 and 47′ each include a plural deflecting electrodes 46 distributed about the axis of symmetry 2. The number of deflecting electrodes 46 can be, for example, two, fours, eight, as in the illustrated embodiment, or even more than eight. The deflecting electrodes 46 are connected to the controller 42 via connectors 48 such that the controller 42 can adjust angles and orientations of deflections provided by the deflectors 47, 47′ to the ion beam 11. Similarly, the detector 21 is connected to the controller 42 via a connector 53 such that the controller can receive detection signals produced by the detector 21. The controller 42 may have a configuration similar to that illustrated with reference to FIG. 3 above. In particular, the controller 42 may include an interface for receiving parameters of a task to be performed by the ion beam system 1, a calculation module configured to determine commands and command parameters suitable for controlling the beam blanker 55, the first and second deflectors 47, 47′ and an acquisition of measurement data via detector 21. The controller 42 may further include a command generation module for encoding the calculated commands and command parameters into digital command data, and a network for distributing the digital command data to control modules controlling the components of the ion beam system 1, such as a beam blanking module, a beam deflection module for the first beam deflector 47, a beam deflection module for the second beam deflector 47′ and a data acquisition module for acquiring the measurement data from the detector 21. FIG. 6 is a flow diagram illustrating a method of controlling the various modules of the ion beam system 1. In a step 201, parameters of a primary task are determined. In the illustrated example, the task is to scan the focused ion beam along a straight line on the sample and to record corresponding detection signals generated by the detector 21. The corresponding primary task is to control the first and second deflectors such that a voltage applied to opposite deflections electrodes 46 of the first deflector 47 starts to rise from a given voltage level (position) at a time Tss1 with a certain rate (speed), and stops rising a at a time Tse1. A corresponding procedure is defined for the second deflector 47′, wherein the positions and speeds may vary between the deflectors since the second deflector 47′ has to deflect the beam by a larger amount than the first deflector 47. Also the start times and end times will vary between the deflectors due to a traveling time of the ions between the first deflector 47 and the second deflector 47′. The parameters of the primary task can be entered by the user, or they can be generated by some software and supplied to the controller 42 via its interface. The parameters of the primary task can be represented by a data structure which is shown in FIG. 6 as a table 203, in which separate lines indicate separate commands for operating components and in which the columns represent the time when a given command is to be executed, a component performing the command, a command performed by the component and parameters of the respective command. In a step 205, additional secondary tasks are determined which are used to perform the primary task. For example, in order to start deflection with deflector 1 using a given voltage level (position) at time Tss1, the initial voltage level (position) is set at an earlier time Tss1−Δ, wherein Δ is selected such that the voltage applied to the pair of electrodes 46 has settled and is sufficiently stable at time Tss1. A similar procedure is applied to the second deflector. Moreover, the beam blanker is controlled to un-blank the beam at a time Tub which is earlier than the start of the deflection by deflector 1 due to traveling times of the ions between the beam blanker and the first deflector. A time Tb for the beam blanker to blank the beam is also determined. Moreover, commands for starting the data acquisition at a time Tas and for terminating the data acquisition at a time Tae is determined. The time Tas will be later than the time Tss1 for starting the deflection with the first deflector due to traveling times of the ions towards the sample and of secondary particles from the sample towards the detector 21. A data structure representing the commands after the secondary tasks have been added can be represented as a table 207, in which different lines represent different commands and columns represent times, components, commands and parameters. The step 205 of adding secondary tasks can be performed by the calculation module of the controller 42. Certain commands of the table 207 can be combined into combined commands in a step 209. For example, the commands of instructing the first deflector to start deflecting at Tss1 and to stop deflecting at Tse1 can be combined to a combined command which instructs the first deflector to start deflection at Tss1 wherein a duration of the deflection is a parameter of the command. Similarly, the commands for starting and stopping the deflection of the second deflector and the commands for un-blanking and blanking the beam can be combined into combined commands having a duration as a parameter. A data structure representing the combined commands can be represented as a table 211. The step 209 of combining tasks can also be performed by the calculation module of the controller 42. In a step 213, some of the commands are corrected for delays by control modules and electronic components and circuits between receipt of the respective commands and start of execution of the commands. For example, the time Tss1 for starting the deflection with the first deflector is corrected by a delay δtd by the deflection module to receive and analyze the command and to set electronic circuits such as voltage generators in order to perform the deflection. The corrected time Tcss1=Tss1−δd indicates the command time at which the command is to be sent to the network such that the beam scanning starts at the time Tss1. Similarly, The command time Tcub for instructing the beam blanker to un-blank the beam at Tub is determined by subtracting a delay δtb from Tub, and the command times Tcas and Tcae of the commands instructing the data acquisition module to start and end data acquisition are determined by subtracting a delay δta from Tas and Tae, respectively. In the illustrated example, the times Tss1−Δ and Tss2−Δ are not corrected for additional electronic delays since the time Δ has been selected such that the initial voltages are set sufficiently ahead of the times Tss1 and Tss2, respectively. The command times of these commands for instructing the beam deflection module are equal to Tss1−Δ and Tss2−Δ, accordingly. A data structure representing the commands generated in step 213 is shown as a table 215 in FIG. 6. Also the step 213 can be performed by the calculation module of the controller 42. The commands generated in step 213 are sorted by command time in a step 217 such that the commands are arranged according to their corresponding times. The sorted commands are shown as a table 219 in FIG. 6. Also the step 217 can be performed by the calculation module of the controller 42. Thereafter, the commands are encoded into digital command data associated with a command schedule in a step 221. The digital command data of each command can be sent across the network to the receiving control modules of the controller 42. The digital command data and schedule are represented as a table 223 in FIG. 6. The encoding of the commands in step 221 can be performed by a command generating module of the controller 42 after having received the calculated commands from the calculation module. The command generation module will then send the sequence of digital command data to the network at times defined by the schedule in a step 225. A start time of sending the sequence of commands can be defined by a trigger signal generated by a clock or supplied separately. The digital command data are received by the beam blanking module, the beam deflection modules and the data acquisition module from the network, and the modules interpret the commands and control the beam blanker, deflectors and data acquisition components such that the commands are executed as desired. In the example illustrated with reference to FIG. 6 above, each digital command includes an instruction for one module of the charged particle system. According to other examples, the digital commands can be generated such that some or all digital commands each include instructions for more than one module, as illustrated with reference to FIG. 4b above. In the examples illustrated above, the method of operating a particle beam system using sorted commands sent to a network such that they are received by a beam deflection module and a beam blanking module are used for performing a line scan with the charged particle beam. However, these methods can also be used to perform other procedures with the charged particle system, such as modifying a sample by deposition of material on or removal of material from a sample, which may be assisted by supplying a reactive gas to the sample, or writing a pattern into a resist. These procedures involve an operation of deflecting the beam to a target position on the sample and, when the beam has reached the target position, un-blanking of the beam such that a dose of charged particles is delivered to the surface of the sample in order to perform the action, such as removal of material, deposition of material and modifying a resist. Depending on a configuration of the pattern, initial deflections to deflect the beam such that it is directed to a target position within a particular pattern feature depends on a distance of this particular pattern feature from another pattern feature which was previously processed. For example, to move the beam from a previously processed pattern feature to a closely adjacent pattern feature involves a relatively small deflection, whereas moving the beam from a previously processed pattern feature to a distant next pattern feature will involve a substantially larger amount of deflection. Depending on the deflection amount, different settling times will be used after completion of the deflection until the beam is stable and points to the desired target location within the next feature. Generally, such settling times or additional waiting times are greater for greater deflections. In order to perform the desired action, such as writing a pattern, with a high accuracy, the beam is un-blanked only after such settling time has expired. With the methods illustrated above, such additional and variable waiting times can be easily achieved. According to an exemplary embodiment in this context, the beam un-blank time is later than the deflection stop time, and a time difference between the beam un-blank time and the deflection stop time is variable. For example, this time difference can be varied by more than 5 μs and 50 μs. In the embodiments illustrated above, one particle beam column, such as an electron beam column shown in FIG. 1 and an ion beam column shown in FIG. 5 are controlled by using methods for generating digital commands for controlling modules of the particle beam column as illustrated above. However, it is also possible to control modules distributed across plural particle beam columns using such methods. For example, a system including an electron beam column and an ion beam column can be controlled with such methods, wherein various modules, such as deflectors, beam blankers and detectors, of the two particle beam columns are connected to a common data network. The modules of the two particle beams may receive digital commands generated based on sorted data records including commands representing instructions for the individual modules. The digital commands can be generated by one controller, for example. While the disclosure has been described with respect to certain exemplary embodiments thereof, it is evident that many alternatives, modifications and variations will be apparent to those skilled in the art. Accordingly, the exemplary embodiments of the disclosure set forth herein are intended to be illustrative and not limiting in any way. Various changes may be made without departing from the spirit and scope of the present disclosure as defined in the following claims. |
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044407190 | summary | BACKGROUND This invention relates to the steam-driven injection of water into a vessel and more particularly relates to the cooling of water subject to multiple-stage steam-driven injection into the pressure vessel of a nuclear system. Pressure vessels in nuclear reactors may operate at pressures in excess of 1,000 psi in the course of generating steam for power production. It is desirable to employ the steam generated within the pressure vessel to insure the presence of an ample supply of cooling water in the core of the nuclear system. Presently, as for example shown in U.S. Pat. No. 3,431,168 (which relates to an invention by J. E. Kjemtrup, was issued on Mar. 4, 1969, and is expressly referred to and incorporated herein, a sufficient water level within the pressure vessel is insured by employing a turbine-driven pump which is an expensive and complex assembly of components including a steam-driven turbine rotating a shaft for driving the pump. In other arrangements, the pump is driven by an electric motor which requires substantial active controls and depends upon electrically operated speed governing equipment for operation. The pump is accordingly subject to failure during a loss of electric power. In the past, and as shown in U.S. Pat. No. 722,696, issued to W. W. Green on Mar. 17, 1903, a form of multiple-stage steam-driven injection of water into vessels has been accomplished. However, in such kinds of steam-driven water injection, irrespective of whether the stages are arranged in series or parallel or combinations thereof, the output of each stage produces heated water, which reduces the efficiency of subsequent stages. OBJECTS OF THE INVENTION Accordingly, it is an object of this invention to provide a passive, or non-electrically operated multi-stage steam-driven water injection system for supplying water to the pressure vessel of a nuclear system that is not subject to electrical failure, since it can be selectively and manually controlled. Another object of the instant invention is to cool the water departing from selected steam-driven stages of a multiple-stage water injection system. It is a further object of the present invention to supply water to the pressure vessel of a nuclear system by employing an injection or water supply system that is inexpensive, uncomplicated, and simple to maintain. SUMMARY The invention calls for stepwise increasing the pressure level of cooling water available from one or more of various containers such as for example the fuel pool, the suppression pool, or the condensate tank of a nuclear system. The increased pressure level is required to overcome the back pressure of the pressure vessel into which the water is sought to be injected and which operates at pressures of about 1,000 psi or above. The desired output pressure level of the injection system is attained in stages by employing steam generated in the pressure vessel itself. Two steam-driven water injectors are arranged in series with a cooling device such as for example a heat exchanger cooling the water flowing therebetween. A jet pump initially receives the water from one or more of the containers to which reference has already been made. Feedback water from the output of the first steam-driven water injector is cooled and propels the input water through the jet pump and into the first steam injector. |
053295633 | summary | BACKGROUND OK THE INVENTION 1. Field of the Invention The present invention relates generally to controls rods of a nuclear reactor. More specifically, the present invention relates to a control rod latch tool which facilitates the release and removal of the control rods when replacement is required. 2. Description of the Relevant Art FIG. 1 is a perspective partially cut-away view of a boiling water reactor of the type to which the present invention is applicable. This BWR (boiling water reactor) includes, merely by way of example, a head 40 which is secured to the top of the reactor vessel 41, a vent and head spray 42, a steam dryer lifting lug 43, a steam drier assembly 44, a steam outlet 45, a steam separator assembly 46, a feedwater inlet 47, a core spray inlet 48, a feedwater sparger 49, a low pressure coolant injection inlet 50, a core spray line 51, a core sprayer sparger 52, a top guide 53, a jet pump assembly 54, a core shroud 55, fuel assemblies 56, a jet pump/recirculation water inlet 57, a core plate 58, control rods 59 (enclosed in elongate tubular guide thimbles), a recirculation water outlet 60, a vessel support skirt 61, a shield wall 62, control rod drive mechanisms 63, control rod drive hydraulic lines 64, and an in-core flux monitor 65. In this type of BWR the control rods 59 have an essentially cruciform cross-section, and, in order to facilitate refueling (viz., replacement of the fuel assemblies), are disposed at the bottom of the reactor vessel in a manner wherein they can be selectively moved up between the fuel rod assemblies by the hydraulically operated locking piston type control rod drive mechanisms 63. FIG. 2 shows an example of a control rod 59 which is used in this type of reactor. This illustrated arrangement includes a lifting handle 66, upper guide rollers 67, a sheath 68, a plurality of neutron absorbing rods 69 enclosed in the sheath to define four blades 70, a coupling release or unlatching handle 71, a velocity limiter 72, lower guide rollers 73, and a coupling socket 74. Details of the construction of the lower portion of a fuel rod is shown in FIGS. 3 and 4. As shown, the coupling socket 74 houses a lock plug 75 and a lock plug return spring 76 which are arranged at the lower end of an actuating shaft 77. When coupled to a control rod drive, the coupling socket 74 is arranged to operatively engage with a spud 78, and an unlocking tube 79 which extend up from an index tube--drive 80, in the manner illustrated in FIG. 4. In order to disconnect a control rod 59 of the above described nature from an associated control rod drive, it is necessary to raise the unlatching handle 71. This releases the coupling arrangement shown in FIG. 4 and allows the control rod 59 to be lifted by the lifting handle 66 located at the top of the rod, up and out of the reactor vessel 41. As is well known, when a reactor of the illustrated type is opened for refueling or maintenance, the vessel 41 is flooded with water in order to suppress radioactive emissions. This, in combination with the fact that the control rods 59 are located deep in the reactor vessel 41, as will be appreciated from FIG. 1, renders access very difficult. Nevertheless, it is required to be able to periodically remove the control rods and/or the control rod drives from the reactor vessel. FIGS. 5 to 10 show an existing latch tool which is used to lift the unlatching handle and release the control rod 59 from its respective control rod drive. This tool basically comprises a stud 81 which extends upward from a frame 82 for connection to a crane line, a tube assembly 83, an actuator link air cylinder 84, a lifting hook air cylinder 85, and an actuator link 86 which is adapted to engage the unlatching handle 71. The actuator link 86 is operatively connected with the air cylinder 84 by way of an actuator rod 87. As shown, the actuator link 86 is supported near the bottom of a V-cross-section structural member 88. Details of a lifting hook structure 89 and the manner in which it is pivotally mounted and connected with the lifting hook air cylinder 85 are illustrated in FIG. 6. In order that the actuator link 86 be accurately located, the arrangement is provided with a short blade-shaped back guide member 90. Accordingly, when this tool 100 is put into use, the tool engages the unlatching handle of the control rod 59 from one side while the back guide 90 provides an aligning function on the other. The actuator link air cylinder 84 and the lifting hook cylinder 85 are supplied with air under pressure by way of air connectors 91. However, with this existing type of arrangement it is necessary for the fuel support piece to be removed before suitable disposition and alignment of the tool can be achieved. This removal is quite time consuming. Accordingly, it has been proposed to remove the back guide 90, from the latch tool in order enable its use without the need to remove the fuel support piece. However, this measure has not met with success in that, without the back guide tool engagement is unreliable and results in problems with the engagement to the unlatching handle. An alternate technique which has been proposed in order to save time and avoid the handling problems related with the removal of the fuel support piece, is such as to use a J-hook which is suspended from a cable, to unlatch the control rod 59. However, this technique suffers from the drawbacks that, as the unlatching handle is submerged in approximately 60 feet of water, in order for the operator to be able to observe the hook and control the maneuvering of the same, it is necessary to utilize a TV camera and immerse it to essentially the same depth as the hook. Correct positioning of both the hook and the camera is time consuming, requires at least two operators and requires a certain amount of experience and skill. SUMMARY OF THE INVENTION It is an object of the present invention to provide a new control rod latch tool which simplifies the unlatching of a control rod from its associated control rod drive. It is a further object of the present invention to provide a control rod latching tool which can be used for both unlatching and lifting the control rods. A further object of the present invention to provide a control rod latch tool which enables the tool to be placed in position on a control rod without the need to remove the fuel support pieces first. In brief, the above objects are achieved by a latch tool which features two elongate arms which extend down along either side of a control rod. Rotatable actuating rods are mounted on the arms. The upper ends of the actuating rods are provided with cranks which engage in slots formed in an actuating disc. As the disc is rotated, the cranks rotate the actuating rods and swing cam-like members, which are fixed to the rods into position under an unlatching handle. In the preferred embodiments, the actuating disc is threaded onto the exterior of a rotatable input member, so that after the cam-like members are swung into position, the actuating disc threads its way up the input member without undergoing any rotation and in a manner which pulls the cam-like members upwardly and lifts the unlatching handle. More specifically, a first aspect of the present invention resides in a tool which is designed to facilitate the unlatching of control rods from a control rod drive, and which features: a rotatable adapter member through which rotational energy can be input; an actuator disc operatively connected with the rotatable member so that rotation of the rotatable adapter member induces rotation of the actuator disc, the actuator disc having an elongate slot formed therein; and an actuator shaft having a crank at a first end and a cam-like member at a second end, the crank being received in the elongate slot so that rotation of the actuator disc induces the actuator shaft to rotate and cause the cam-like member to rotate through a predetermined angle. A second aspect of the present invention resides in a tool for use in a nuclear reactor and which can be lowered onto an essentially cruciform cross-section control rod having a lift handle at the top and an unlatching handle arranged near the bottom, the control rod being releasably connected with a control rod drive by a connection mechanism which is operatively connected with the unlatching handle, the tool comprising: first and second arms which extend down along opposite sides of the control rod from a cross-member which is seatable on the top of the control rod; first and second actuator rods which are supported on the first and second arms; first and second cam-like members which are fixed to the first and second actuator rods, respectively; a rotatable drive input member which is rotatably mounted on the cross member; and crank means operatively interconnecting the input member with the first and second actuating rods, for selectively inducing the first and second actuator rods to rotate and to induce the first and second cam-like members to engage and then lift the unlatching handle. As will be appreciated from the above, features which characterize the present invention reside in that the tool can be placed in an operative position without the need to remove the fuel support piece and when lowered down onto the control rod, becomes symmetrically located on both side of the control rod, in a manner which prevents separation of the control rod and the latch tool. This of course ensures reliable engagement with the unlatching handle. As the tool is used with a so called "rigid pole" system, precise location and operation is facilitated. The engagement, unlatching movement, and disengagement of the tool is easily carried out, without the need for air cylinders and valves, by a single operator who only needs to rotate the rigid pole through a number of turns . |
description | This application is a Divisional application of Ser. No. 10/106,895, filed Mar. 27, 2002, now U.S. Pat. No. 6,836,523 which claims priority from Japanese patent application 2001-093306, filed Mar. 28, 2001, and Japanese patent application 2002-046788, filed Feb. 22, 2002. The entire contents of each of the aforementioned applications are incorporated herein by reference. 1. Field of the Invention This invention relates to a device and a method for radiation measurement, applied to monitor the radiation in an extensive range for improving resistance to noises in a digital signal processing. 2. Description of the Related Art As far as radiation measurement is concerned, if wide range radiation is measured, then the pulse measurement method and the Campbell measurement method are often used together. Generally, the pulse measurement method counts the pulse number of a pulse signal from a radiation sensor, but if the pulses overlap and it cannot count by the pulse measurement method, the Campbell measurement method is performed. For example, from six to ten start-up range neutron monitor sensors (SRNM sensors) and from one hundred to two hundred local power range monitor sensors (LPRM sensors) are installed inside of a reactor pressure vessel containing nuclear reactor core to monitor nuclear reactor power. A start-up range neutron monitor and a power range neutron monitor measure outputs of the SRNM sensors and the LPRM sensors, respectively, to monitor the nuclear reactor power in a monitoring range of about eleven figures. In this composition, the start-up range neutron monitor is used to count the pulse number of an output signal of the SRNM sensor in order to monitor relatively low reactor output, that is, the output is in from 10−9% to 10−4% of effective full power of the reactor. This is henceforth called the pulse measurement method. On the other hand, the Campbell measurement method, that is, the measuring of fluctuation power generated due to overlapping of the pulse outputted from the sensor, is used in order to monitor relatively high reactor output, that is, the output is in from 10−5% to 10% of the effective full power of the reactor. Hereafter, a conventional technical example of the pulse measurement method and the Campbell measurement method in a nuclear reactor start-up monitoring system, which is disclosed in Japanese Patent Disclosure (koukai) No. 2000-162366, which is equivalent to U.S. Pat. No. 6,181,761, is explained with reference to FIG. 18. The nuclear reactor start-up monitoring system shown in FIG. 14 is composed of an SRNM sensor 1 for outputting an electric signal containing pulse components corresponding to the number of neutrons in response to neutrons generated in the nuclear reactor, an analog preamplifier 2, an A/D (analog-to-digital) converter 3, and a pulse counter 23, an integration counter 24, a power calculator 25, an arithmetic average calculator 26, and a reactor power monitoring system 27. The analog preamplifier 2 amplifies the electric signal having pulse components outputted from the SRNM sensor 1 to regularize the electric signal, and the A/D converter 3 converts an analog signal outputted from the preamplifier 2 to digital data sampled at intervals which are shorter than a pulse width of the pulse included in electric signal outputted from the SRNM sensor 1. The pulse counter (PC) 23 counts a number of pulses in the sampled data outputted from the A/D converter 3 and converts the number of the pulse to an output level value contained in relatively low range power of the nuclear reactor, and the integration counter 24 adds the sampled valve outputted from the A/D converter 3 to raise the measurement accuracy. The power calculator 25 calculates a power by squaring the added value of the integration counter 24, and the arithmetic average calculator 26 averages the power calculated by the power calculator 25. The reactor power monitoring system 27 continuously monitors the output at the start-up of the nuclear reactor based on the counter result of the pulse counter 23 and the calculation result of the arithmetic average calculator 26. In the digital reactor start-up monitoring system of such a composition, the preamplifier 2 amplifies and regularizes a shape of a pulse included in the electric signal outputted from SRNM sensor 1, and the A/D converter 3 samples the amplified and regularized pulse at high speed and calculates the pulse by using one or more logical operations. Also, the pulse counter 23 counts the calculation results outputted from the A/D converter 3 as an output pulse of the sensor if each calculation result outputted from the A/D converter 3 is in a corresponding predetermined range, respectively. On the other hand, the same sampled value is added in the integration counter 24 to lower into a level of a sampling rating required for the Campbell measurement method and to earn a dynamic range for improving the number of equivalent bits. The power calculator 25 adds square values of the results after performing band-pass-filter process for the results, and the arithmetic average calculator 26 averages the results calculated by the power calculator 25 and computes the Campbell output value. The pulse enumerated data and the Campbell output value are estimated by the nuclear reactor output evaluation unit 27 and are displayed as a nuclear reactor output. In this composition, calculation limited to the sensor-outputting pulse can be carried out with excluding noises having long pulse widths by discrimination based on information of not only a pulse height of a pulse but a pulse width by the pulse calculator 23. That is, in the reactor start-up monitoring system of FIG. 18, for example, the output signal of the SRNM sensor 1 containing a pulse with the pulse width of 100 nanoseconds is sampled at intervals of 25 nanoseconds. Four sampled-data, from data No. k−3 to data No. k, denoted as S(k−3), S(k−2), S(k−1), and S(k) in order, respectively, which correspond to a pulse width, are used to calculation described below, as S(k−3) is a sampled value at a rise point of a pulse, S(k) is a sampled value at a fall point of the pulse, and two sampled data S(k−1), S(k−2) are in between S(k-3) and S(k). It considers a result Out(k) of this calculation as an index of pulse discrimination, and as a result, the pulse is counted as a neutron pulse if it is in a range of predetermined level.Out(k)={b*S(k−2)+c*S(k−1}−{a*S(k−3)+d*S(k)} (1),where a, b, c and d are non-zero constants. By this calculation, it becomes possible to calculate only signals having almost similar pulse widths as that of the output pulse of the SRNM sensor 1. That is, even if a large surge-like noise becomes overlapped on a signal pulse, it can count measured value exactly by deducting the ground level of the pulse. In addition, by setting two or more indices such as the Out(k) for detecting a case corresponding to such a sensor pulse form as mentioned above and using AND logic among these indices, this discrimination performance can be improved further. Thus, even if a surge-like noise with a pulse width of several microseconds overlaps, and is supposed to be guided into a pulse in an electric signal outputted from the SRNM sensor most easily, the surge-like noise can be removed nearly completely and a limited calculation of sensor pulses with a pulse width of about 100 nanoseconds can be performed. On the other hand, in the Campbell measurement method, the power calculator 25 restricts a frequency band and calculates an average of square values of the sampled data. In this composition, since the frequency band can be set up by software programming, if a noise is in a certain frequency equivalent to a measurement band, changing the measurement band on the software programming can reduce guidance of the noise. However, there are several subjects described below in the nuclear reactor start-up monitoring system according to the above-mentioned conventional technology. A first subject concerns reduction of a bipolar noise. That is, if surge noise with a pulse width of several microseconds and sensor output pulse overlap, it is necessary to compute a value corresponding to a pulse peak value by using the difference between them in order to count the overlapped sensor output pulse without preparing dead time. In taking the difference, if the pulse is homopolar, that is, either a positive pulse or a negative pulse, such as a sensor output pulse, a pulse discrimination level of the pulse is equivalent to a conventional pulse peak value from the ground level. However, if the pulse is bipolar, such as a white noise from a circuit resistance, it is necessary to discriminate voltage between peaks of the pulse from the pulse discrimination level. For this reason, the discrimination level necessary in this case is twice as much as that of conventional discrimination method using pulse peak from ground level. Therefore, the discrimination level required to count the sensor output limitedly is needed about twice as much as that of the conventional method, and thus the ratio of sensor signal to white noise, that is, the signal-to-noise ratio (S/N ratio), worsens. A second subject concerns improvement of resistance to noises in the Campbell measurement method. Conventional noise test of a motor, for example, shows that a surge noise with a pulse width of several microseconds is easily induced to the reactor start-up monitoring system. In the pulse measurement method, this surge noise can be reduced by pulse discrimination according to the above-mentioned digital calculation. On the other hand, in the Campbell measurement method, a measurement band is set as a frequency band from several hundreds of hertz to one megahertz, which is selected according to a form of a sensor output pulse, and in the above-mentioned precedence example, the induction noise is removed by shifting this measurement band. However, since the frequency of the noise that is the easiest to be guided mostly falls in a range of the measurement band, it is difficult to remove the noise completely, and it is necessary to rectify sensor sensitivity because the sensitivity changes slightly. Generally, a measurement device for measuring dose equivalent is optimized in a radiation incidence window, reaction volume, etc., of a sensor, so that sensitivity characteristics over gamma ray energy of the device may become equal to energy absorption characteristics of a human body. However, it is difficult to make the sensitivity characteristics in agreement correct because the sensitivity characteristics differ according to directions of incidence of gamma rays. Moreover, as far as accurate conversion of the dose equivalent to a human body is concerned, since energy absorption characteristics differ according to parts of a human body, it is difficult for independent use of the measurement device modified to equalize to the sensor sensitivity over gamma ray energy to evaluate the dose equivalent in each part of a human body. Furthermore, when neutrons other than of a gamma ray, such as a beta ray, are intermingled, a sensor that has rectified its sensitivity by arranging sensor structure cannot estimate such mingled radiations, each of which has absorption characteristics which are greatly different from that of another radiation. Therefore, it must arrange a plurality of measurement systems each of which is used for measuring one radiation exclusively. Conventionally, in order to solve these subjects, it is proposed and put in practical use to compute energy spectrum of a gamma ray to be converted to the dose equivalent. However, since this technique is based on acquisition of energy information by using pulse height, in a condition in which pileup of pulses is occurred, it becomes difficult to acquire the energy information and thus the accuracy of this technique worsens. That is, although depending on a pulse width of a sensor output pulse, a maximum of conventional energy measurement is about 1*105 counts per second (CPS). If it is supposed that a minimum of the measurement is one CPS, which must satisfy a response demand, a measurement range goes into about 5 figures. Thus, it is desired to realize a measurement method which enables to measure a dosage in more extensive range continuously. Japanese Patent Disclosure (koukai) No. H3-183983 shows that dual structure of sensors in a radiation measurement device for measuring dose equivalent in depth of one centimeter improves measurement precision. In this technique, the above-mentioned pileup influence in the pulse measurement method is evaded by means of measuring current. However, sensor structure and processing in this technique are complicated, thus it is desired that they be simplified. Accordingly, an object of this invention is to provide a device and a method for measuring radiation which improves noise resistance in the pulse measurement method and the Campbell measurement method using digital processing. Another object of this invention is to provide a device and a method of radiation measurement which monitors a dosage in a wide range continuously with a convenient composition by applying the Campbell measurement method to the measurement of a radiation dosage. Additional purposes and advantages of the invention will be apparent to persons skilled in this field from the following description, or may be learned by practice of the invention. According to an aspect of this invention, there is provided a device for measuring radiation, including a radiation detector which generates an analog signal containing pulse components corresponding to a dosage of an inputted radiation, an A/D converter which regularizes the analog signal outputted from the radiation detector and converts the regularized analog signal into sampled data, an n-th power pulse discrimination unit which calculates an n-th power value for each of the sampled data outputted from the A/D converter and discriminates the pulse components contained in the analog signal of the radiation based on the calculated n-th power values to generate a discrimination signal associated with each discriminated pulse component, where n is an integer of not less than two, and a pulse counter which counts a number of the discriminated pulse components based on the discrimination signal outputted from the n-th power discrimination unit. According to another aspect of this invention, there is provided a device for measuring radiation, including a radiation detector which generates an analog signal containing pulse components corresponding to a dosage of an inputted radiation, an A/D converter which regularizes the analog signal outputted from the radiation detector and converts the regularized analog signal into sampled data, a band pass filter which limits the sampled data outputted from the A/D converter within a predetermined frequency band to generate restricted sampled data, an n-th power calculation unit which calculates the n-th power values of the restricted sampled data outputted from the band pass filter, where n is an integer of not less than two, a first smoothing unit which equalizes the n-th power values of the limited sampled data outputted from the n-th power calculation unit within a first time width to generate a first smoothed n-th power value, a data removal equalization unit which evaluates sizes of the first smoothed n-th power values outputted from the first smoothing unit within a second time width, removes a predetermined data removal number of the first smoothed n-th power values based on the evaluation result, and equalizes the first smoothed n-th power values after the removing within the second time width to generate a second smoothed n-th power value, a second smoothing unit which equalizes the equalized n-th power values outputted from the data removal and equalization unit to generate a third smoothed n-th power value, and a converter which converts the second smoothed n-th power value outputted from the second smoothing unit into a radiation intensity of the inputted radiation. According to still another aspect of this invention, there is provided a device for measuring radiation, including a radiation detector which generates an analog signal containing pulse components corresponding to a dosage of an inputted radiation, an n-th moment calculation unit which calculates an average value of the n-th power values of pulse heights within a time width as an n-th moment value based on the analog signal outputted from the radiation detector, where n is an integer of not less than two, and where the pulse heights correspond to the pulse components included in the analog signal, a pulse counter which counts a number of pulse components based on the analog signal outputted from the radiation detector, an average energy calculation unit which calculates an average energy of the radiation based on a ratio of the n-th moment value calculated by the n-th moment calculation unit to the number of the pulse components counted by the pulse counter. According to still another aspect of this invention, there is provided a device for measuring radiation, including a radiation detector which generates an analog signal containing pulse components corresponding to a dosage of an inputted radiation, an n-th moment calculation unit which calculates an average value of the n-th power values of pulse heights within a time width as an n-th moment value based on the analog signal outputted from the radiation detector, where n is an integer of not less than two, and where the pulse heights correspond to the pulse components included in the analog signal, a current measurement instrument which calculates an average current from the pulse heights of the pulse components included in the analog signal, and an average energy calculation unit which calculates an average energy of the radiation based on a ratio of the n-th moment value calculated by the n-th moment calculation unit to the average current calculated by the current measurement instrument. According to still another aspect of this invention, there is provided a device for measuring radiation, including a radiation detector which generates an analog signal containing pulse components corresponding to a dosage of an inputted radiation, first to n-th moment calculation units each calculating an average value of one of first to n-th power values of pulse heights corresponding to the pulse components included in the analog signal within a time width as one of first to n-th moment values, respectively, where n is an integer of not less than three, and an average energy calculation unit which calculates an average energy of the radiation based on a ratio of two of the first to n-th power values calculated by the first to n-th moment calculation units, respectively. According to still another aspect of this invention, there is provided a method of measuring radiation, including A/D converting an analog signal containing pulse components corresponding to a dosage of an inputted radiation outputted from a radiation detector into sampled data, calculating n-th power values of the sampled data, where n is an integer of not less than two, and discriminating the pulse components of the radiation contained in the analog signal based on the n-th power values of the sampled data. According to still another aspect of this invention, there is provided a method of measuring radiation, including A/D converting an analog signal containing pulse components corresponding to a dosage of an inputted radiation outputted from a radiation detector into sampled data, calculating n-th power values of the sampled data, where n is an integer of not less than two, equalizing the n-th power values of the sampled data within a time width; and discriminating the pulse components of the radiation contained in the analog signal based on the equalized n-th power values of the sampled data. According to still another aspect of this invention, there is provided a method of measuring radiation, including calculating an average value of n-th power values of pulse heights of pulse components corresponding to a dosage of an inputted radiation included in an analog signal outputted from a radiation detector within a time width, where n is an integer of not less than two; and calculating at least one of a radiation intensity of the inputted radiation and a dosage equivalent of the inputted radiation based on the average value. Referring now to the drawings, wherein like reference numerals designate identical or corresponding parts throughout the several views, the embodiments of this invention will be described below. First Embodiment A radiation measurement device of a first embodiment in this invention is explained with reference to FIG. 1. The radiation measurement device shown in FIG. 1 is composed of an SRNM sensor 1 for generating an electric signal containing pulse components according to a radiation dosage in response to an inputted radiation, a preamplifier 2A for amplifying the output pulse, an A/D converter 3 for sampling the output pulse of the preamplifier 2A at intervals of time shorter than pulse duration of the output pulse to obtain sampled data, an n-th power pulse discrimination unit 4, and a pulse counter 5. The n-th power discrimination unit 4 is provided to calculate an n-th power value of the sampled data, corresponding to the pulse duration of the pulse from the SRNM sensor, and to discriminate a signal by comparing the calculated n-th power valve with a predetermined discrimination level. And the pulse counter 5 counts a pulse discriminated by the n-th power discrimination unit 4. The SRNM sensor 1 is a nuclear fission sensor for outputting a signal containing pulse components, and it can also replace the sensor by an ionization chamber from which same kind of the pulse output of the SRNM sensor is obtained. In such a composition, when neutrons are injected into the SRNM sensor 1 of this radiation measurement device and nuclear fission is occurred in the sensor 1, an electric analog signal containing pulse components as shown in FIG. 2(a) is outputted from the SRNM sensor. A pulse width of the pulse in this signal outputted from the SRNM sensor 1 is about 100 nanoseconds. This output signal is inputted into the preamplifier 2A and the pulse is amplified. The preamplifier 2A also has a function to impress an operating voltage to the SRNM sensor 1. The signal with pulse components outputted from the preamplifier 2A is inputted into the A/D converter 3, and is sampled at sampling time intervals to be digitalized, as sampled data are shown by dots in FIG. 2(a). The shorter these sampling time intervals are, the more information about waveforms can be extracted, and if these sampling time intervals are sufficiently short, it is possible to count only the output pulse of the sensor correctly with excluding a signal due to an incoming foreign noise. The A/D converter 3 also performs band-pass-filter processing for restricting to a frequency band which is necessary in the sampling theorem before the sampling of data. The sampled data outputted from the A/D converter 3 is inputted into an n-th power calculation unit in the n-th power pulse discrimination unit 4 to calculate the n-th power value of the data. That is, the n-th power calculation unit calculates an n-th power value of each sampled data, or multiplies by n pieces of sampled data which are placed sequentially. Here, n is an integer of not less than two. In case of calculating the n-th power value of each sampled data, it also performs equalization processing of two n-th power values placed sequentially. For example, when a pulse waveform is sampled at eight pieces, it performs equalization processing of two pieces of data placed sequentially after the calculating of each square value of each data, and consequently four values are acquired. In this case, it is also possible to perform moving-average processing to acquire eight pieces of sampled data. FIG. 2(b) shows a trend of sampled data when performing square value calculation in a case that n is two, as one example. By transforming each sampled data into a square value, a pulse height ratio of the sensor output pulse to a noise component due to the circuit resistance can be improved to n-th power times as much as that of the conventional method. However, when it uses values acquired by simple calculation of the n-th power value of every sampled data for pulse discrimination, the discrimination performance is the same as that in a case the sampled data is used for the pulse discrimination without calculation. Then, when calculating of the n-th power value in the digital operation, it is surely necessary to add a processing of multiplication of several pieces of data placed sequentially or a processing of equalization of the several pieces of data placed sequentially, after the n-th power calculation, as already stated. By comparing the calculation result acquired by this square calculation with a predetermined discrimination level, which has a minimum and a maximum, and recognizing the result comes from a sensor output pulse when the calculation result is within the predetermined discrimination level, it becomes easier to discriminate an output pulse of the SRNM sensor 1 from a coming foreign circuit noise. By this radiation measurement device, a pulse discriminated by an output of the SRNM sensor 1 is converted into a pulse generating rate in the pulse counter 6, and is finally converted to a neutron flex level in a position of the SRNM sensor 1. According to this embodiment of the invention, it can discriminate a pulse which has a pulse height of the same grade as a circuit noise level better than the conventional method calculating a difference. FIG. 3(a) shows an example of sampling of an amplified electric signal containing a white noise, which is one of foreign noise due to a circuit resistance with a relatively short pulse width, and a sensor output pulse outputted from the SRNM sensor 1 with a pulse height of the same grade as that of the white noise by A/D converter 3. FIG. 3(a) shows a case that a sensor output pulse is generated at around 4.20*103 nanoseconds, and in a section between two vertical dashed lines the sensor output pulse is overlapped with the white noise. And in an area other than this section, there is no sensor output pulse. Suppose that a horizontal dashed line Lb in FIG. 3(a) denotes a maximum of a discrimination level, if the conventional noise discrimination by a pulse height is performed to these data, one pulse at around 4.20*103 nanoseconds including a white noise and a sensor output pulse is included in the discrimination level and another one pulse-like portion at around 4.05*103 nanoseconds including a white noise and no sensor output pulse is also included in the discrimination level, thus the pulse counter counts both the pulse and the pulse-like portion. That is, as a result, a portion in which no output signal occurs is also counted, so in this method it cannot carry out an exact measurement. Moreover, in the above-mentioned pulse count method using a difference between sampled data values, if a circuit noise is generated as a bipolar noise with both positive and negative components, a voltage difference between a positive peak and a negative peak of the bipolar noise in a certain time width is recognized as a pulse height in this time width. Thus, it cannot discriminate the circuit noise unless it raises a discrimination level to twice the voltage as that in the conventional discrimination method by seeking a pulse height value from zero volt. On the other hand, FIG. 3(b) shows a calculation result of the sampled data shown in FIG. 3(a) acquired by a conventional method for calculating an arithmetical mean among three values lined sequentially. By calculating an arithmetic average, positive and negative components of the bipolar noise are cancelled and equalized. However, as far as a homopolar sensor output signal is concerned, it has originally one of a positive component and a negative component; thus the above-mentioned cancellation cannot be cancelled and the pulse width of the calculation result of data of such a homopolar signal becomes longer as shown in FIG. 3(b). While the pulse width becomes longer, in a condition in which there are a lot of output pulses included in a signal outputted from the sensor, there is a possibility where the pulses may overlap and the counting of a number of the pulses cannot be performed correctly and an upper count limit of the pulse measurement becomes lowered. FIG. 3(c) shows a calculation result of the sampled data shown in FIG. 3(a) in the n-th power pulse discrimination unit 4 in this embodiment, here, for example, by calculating the square values of the sampled data values and afterward calculating an arithmetic average of three sequentially-lined square values. By comparing a surrounded portion of two vertical dashed lines, including both the white noise and the sensor output pulse, with another portion left of the surrounded portion, it is found that calculated values in the surrounded portion definitely differ from calculated values in the another portion, thus the discrimination can be performed. That is, by setting a minimum of the discrimination level relatively close to zero, for example, around 1*104 volt**2, the pulse counter can count only the sensor output pulse, therefore, in this method, the discrimination performance can be improved from the conventional method calculating differences. Moreover, compared with the conventional method as shown in FIG. 3(b), after the calculation, a pulse width composed of the calculated values is not prolonged comparatively; therefore, it can measure the pluses without worsening the upper count limit of the pulse. If n is an odd integer in this embodiment, the above-mentioned method can equalize while maintaining signs of the bipolar noise; therefore and a homopolar signal can be discriminated from a bipolar noise signal with a good signal-to-noise ratio. Therefore, in this embodiment, in setting a discrimination level of an output pulse, the discrimination level for removing a circuit noise or an alpha ray noise is set relatively low, and accordingly, even if the sensor output pulse is small, the pulse can be measured without lowering the measure sensitivity. Second Embodiment A second embodiment according to this invention is explained with reference to FIG. 4. In this embodiment, an n-th power pulse discrimination unit 4 of the radiation measurement device shown in FIG. 4 is composed of an integration discrimination unit 6, a difference discrimination unit 6, and a pulse height and power discrimination unit 8. An output of A/D converter 3 composed of the sampled data values is inputted into the integration discrimination unit 6, a difference discrimination unit 7, and outputs of these units 6, 7 are imputed to a pulse height and power discrimination unit 8. An output signal of the pulse height and power discrimination unit 8 is inputted into a pulse counter 5. Here, in the integration discrimination unit 6, the pulse is discriminated according to the pulse discrimination method as explained in the first embodiment. That is, the integration discrimination unit 6 of the n-th power discrimination unit calculates the n-th power value of the sampled data values, and judges whether there is a sensor output pulse or not by comparing the n-th power values or arithmetic averages of every sequential n-th power values with a predetermined discrimination level. A first example of discrimination method of the difference discrimination unit 7 in this embodiment is explained according to the following principle. Suppose that the maximum value and the bottom value of an output of the A/D converter 3 are denoted as Top(k) and Bottom(k), respectively, namely:Top(k)=b*S(k−2)+c*S(k−1),Bottom(k)=a*S(k−3)+d*S(k),where a, b, c and d are non-zero constants. Then, the pulse height value High(k) in the above-mentioned conventional formula (1) can be denoted to be simplified as:High(k)=+Top(k)−Bottom(k). In this example of this embodiment, firstly, it calculates a difference of a square value of a top value Top(k) and a square value of a bottom value Bottom (k), which is hereafter denoted as X, is calculated, namely:X=+Top(k)2−Bottom(k)2=(Top(k)−Bottom(k))*(Top(k)+Bottom(k))=High(k)*(Top(k)+Bottom(k)) (2) Here, when a sensor output signal pulse is superimposed only on a usual circuit noise such as a white noise (hereinafter it is called Case 1), Top(k) is extremely larger than Bottom(k); therefore the formula (2) can be replaced to an approximate formula such as:X=High(k)*Top(k) (3). On the other hand, when the signal pulse is superimposed on an extremely large surge noise (hereinafter called Case 2), Top(k) equals approximately Bottom(k) as an approximation; thus, the formula (2) can be expressed with an approximate formula such as:X=High (k)*(2*Top(k)) (4). Thus, from the formulas (3) and (4), it holds the relation as:X/Top(k)=a*High(k) (5), Provided a is either one or two, that is, a equals one in Case 1 and α equals two in Case 2; therefore, the value X/Top(k) mostly serves as a linear function of the pulse height High (k). That is, even when the output pulse of the SRNM sensor is overlapped on the surge noise, it becomes possible to discriminate and calculate the SRNM sensor output pulse of several hundreds of nanoseconds which overlapped on the surge noise with a cycle of several microseconds, by the discrimination comparing the above-mentioned value X divided by Top(k) with a predetermined discrimination level. Thus, according to this first example of the second embodiment, even if a foreign noise with a pulse width longer than that of the sensor pulse is induced, the influence due to the foreign noise can be reduced by the discrimination using difference of the n-th power values. Next, a second example of discrimination method of the difference discrimination unit 7 in this embodiment is explained according to the following principle. In the formula (1), say,D1(k)=c*S(k−1)−d*S(k) (6),D2(k)=−a*S(k−3)+b*S(k−2) (7). Thus, the peak value High (k) is denoted as follows:High(k)=+D1(k)+D2(k) (8). The sum of the n-th power values of each member in the right side of the equation (8), denoted as Y hereinafter, isY=D1(k)n+D2(k)n. And this formula is deformed, as an approximation, to the following:Y=High(k)n. That is,Y−n=High (k). Thus, in this case, the pulse discrimination is possible by calculating the value Y− n as an approximate index for comparing with a predetermined discrimination level. In addition, it is equivalent to a formula (1) when n=1 in this case. As mentioned above, even when the output pulse of the SRNM sensor 1 is overlapped on a surge-like noise, it is possible to discriminate and calculate the output pulse of the SRNM sensor 1 for several hundreds of nanoseconds which overlapped on the surge noise with a cycle of several microseconds by using the index acquired by calculating the difference. As mentioned in the first embodiment, for counting a number of pulses, the integration discrimination unit 6 is effective in excluding influence of bipolar noises, such as a white noise, having an incoming interval shorter than the pulse duration of the sensor output, and is also effective in excluding influence of noises with pulse components having pulse heights smaller than that of the sensor output pulse, for example, a circuit noise or an alpha ray noise of the sensor. On the other hand, the difference discrimination unit 7 is effective to noises having a cycle longer than pulse duration of an output pulse of the SRNM sensor 1, and in general it is also effective to remove foreign induced noises having a pulse duration of several microseconds. Therefore, by adjusting logics of these units most suitable, respectively, it can calculate the sensor output only by counting the pulse only when conditions of these units are both effected. As mentioned above, according to the first example and the second example of this second embodiment, by using an n-th power value of a difference of the sampled data corresponding to a pulse height, the influence due to foreign noises with a pulse width longer than that of the sensor output pulse can be reduced, and thus it becomes possible to perform radiation measurement with higher accuracy, as well as the first example of the second embodiment. Next, a composition of the pulse height and power discrimination unit 8 is explained as a third example of this embodiment. The pulse height and power discrimination unit 8 receives an integral value of a pulse from the integration discrimination unit 6 and a value corresponding to a pulse height value of a pulse from the difference discrimination unit 7. A ratio of these values, that is, an integral value divided by the pulse height value, is mostly shown as a certain fixed value equivalent to a pulse width when the pulse is a sensor output pulse. On the other hand, since the white noise containing a high frequency component has a small integration value even if the pulse height value of the noise is equivalent to a sensor output pulse, this ratio of the white noise becomes small. Moreover, since the surge noise with a long pulse width has a large integration value and a small pulse height value, this ratio of the surge noise becomes larger than that of the sensor output. Therefore, by calculating this ratio in the pulse height and power discrimination unit 8 and setting the pulse counter 5 for counting as a pulse when this ratio is within a predetermined certain range, the influence due to these noises can be reduced. As mentioned above, in this pulse measurement method, even if a surge-like foreign noise is induced, when the surge-like noise has a cycle of several microseconds, which is longer than pulse duration of the sensor output pulse, that is 100 nanoseconds, the influence due to the surge-like noise can be eliminated and it can also count pulses overlapped on the noise. Thus, according to this third example of this embodiment, by using both the pulse calculation method using the difference mentioned in the first or the second example of this embodiment and the pulse calculation method using the n-th power value mentioned in the first embodiment, it can measure pulses accurately with accompanying characteristics of the both methods. Third Embodiment Next, a third embodiment in this invention is explained with reference to FIG. 6. A radiation measurement device of the third embodiment shown in FIG. 6 is characterized as a homopolar conversion unit 9 for converting a bipolar signal into a homopolar signal, which is either a non-negative signal or a non-positive signal, according to the polarity of a main component of a pulse contained in an inputted signal, and which is arranged between the A/D converter 3 and the n-th power pulse discrimination unit 4 in the first embodiment of this invention shown in FIG. 1. In addition, like the first embodiment, although the SRNM detector 1 is a nuclear fission detector from which a pulse output is acquired, radiation detectors, such as an ionization chamber from which the other pulse outputs can be acquired, can be applied to the detector instead of the SRNM detector 1 in this embodiment. In this embodiment of such a constitution, a function of the homopolar conversion unit 9 is explained with reference to FIG. 7. FIG. 7(a) shows an example of a pulse waveform when performing secondary differentiation processing to a sensor output, such as a pulse shown in FIG. 2(a). In this embodiment, it is possible to construct the secondary differentiation processing by processing in an analog circuit in the preamplifier 2, or by digitally processing the secondary differentiation calculation to data sampled by the A/D converter 3, both of which are available. A result of average processing after n-th power calculation of the waveform shown in FIG. 7(a), where n is an even number, such as two, by the n-th power pulse discrimination unit 4 is shown in FIG. 7(b). In this calculation, since the bipolar waveform is changed to a homopolar (non-negative) waveform by setting n as an even number, the pulse duration is prolonged. Then, since the main component of the pulse shown in FIG. 7(a) is negative, where the lower direction means negative in FIG. 7(a), the homopolar conversion unit 9 of this embodiment replaces the positive component of the signal shown in FIG. 7(a) with zero or a certain negative value close to zero, and converts to the waveform as shown in FIG. 7(c). A result of average processing after n-th power calculation of this waveform shown in FIG. 7(c), where n is an even number, such as two, is shown in FIG. 7(d). It is possible to narrow a spread of the pulse width compared with a case of FIG. 7(b) without using this homopolar conversion unit 9. Therefore, in this embodiment, even when n is an even number, the pulse width is not prolonged, and thus it is possible to reduce incorrect counting due to pulse pileup which blocks to count one pulse by overlapping of pulses, and prevent a reduction in the measurement minimum of the pulse measurement. Fourth Embodiment Next, a fourth embodiment of this invention is explained with reference to FIG. 8. In this embodiment, a preamplifier 2B having a band restriction function amplifies a signal outputted from the SRNM sensor and restricts the output in a certain frequency band, and then outputs a signal to the A/D converter 3. Digital data outputted from the A/D converter 3 are inputted into a band pass filter (BPF) 10, and is restricted to a specific frequency band. Usually, the frequency band of the band pass filter 10 is settled by output characteristics of the SRNM sensor 1, and for example, the band pass filter 10 can be constituted as a digital filter which passes only the frequency component in the range between 100 kHz and 400 kHz. Although several measurement bands in the band pass filter 9 can be settled, in an explanation hereinafter, it is represented with a composition specifying one band. In the digital filter processing, a sampling period and a number of bits of inputted sampled data are adjusted suitably according to a frequency band of the output of the band pass filter 10. The output of this band pass filter 10 is inputted into an n-th power value calculation unit 11 and is converted into an n-th moment value by calculating an n-th power value of each sampled data value. In this embodiment, in an output side of the n-th power value calculation unit 10, a first smoothing unit 12A, a data removal equalization unit (DEA) 13 and a second smoothing unit 12B are arranged. The first smoothing unit 12A calculates an average of the n-th power values, that is, n-th moment values, within a first time width to output a first smoothed value. The data removal equalization unit 13 removes some of the first smoothed values outputted from the first smoothing unit 12A within a second smoothing time width and afterward calculates an average of the first smoothed values to output a second smoothed value. The second smoothing unit 12B calculates an average of the second smoothed values outputted from the data removal equalization unit 13 within a third smoothing time width to output a third smoothed value. Here, by adjusting a number of the data for calculating an average in at least one of the first smoothing unit 12A, the data removal equalization unit 13 and the second smoothing unit 12B, or at least one of the first time width, the second time width and the third time width, and a number of removing data in each second time width in the data removal equalization unit 13, based on a pulse width and an arrival interval of a surge-like noise and an arrival cycle of a foreign noise, the MSV measurement can be performed without influence from the foreign noises. Still more reliable measurement is realizable in this embodiment. Furthermore, as one deformed example of this embodiment, it is also possible that the first smoothing unit 12A selects a maximum value of the n-th power values within the first time width instead of calculating the average, and afterward the data equalization unit 13 calculates an average of the maximum values within the second time width. By this example, the MSV indicated value can be obtained even if the counting rate is low, and thus a measurement minimum in the MSV measurement decided by a circuit noise of the preamplifier 2B, etc., can be expanded. In addition, if n is set as an odd integer of not less than three in this embodiment, since a surge-like noise is a bipolar noise, the equalization processing offsets positive values and negative values of the noise. And by choosing a value of the same polarity as that of the homopolar sensor output, the influence of the surge-like noise can be reduced. Hereinafter, in this embodiment, a square value represents as the simplest example on the n-th moment operation. That is, it explains by supposing that n equals two and a k-th output sampled data is denoted as S(k) and the n-th power calculation unit 11 calculates:Out1(k)=S(k)*S(k) (10). The outputs Out1(k) are inputted into the first smoothing unit 12A, and a certain number of the outputs are equalized. In this present circumstance, if an average of the output values of the band pass filter 10 has an offset, the average of sampled data S(k) and the square value of the average are also calculated, and it subtracts the square value of the average from the square operation result Out1(k). That is, in equalizing n pieces of values, it calculates:Out2(k2)=(Out1(k))/n−{(Σs(k))/n}2 (11). Here, the sigma Σ shows adding n pieces of sampled data. In this situation, the adding number n is set as a number equivalent to pulse width of an assumed foreign noise, and moreover, this number is arranged as an n-th power of 2 so that it makes easier to carry out digital operation. That is, suppose that the noise is like a noise as shown in FIG. 8, that is, a pulse-like noise including surge-like noises with pulse width of 20 microseconds coming in at intervals of 2 milliseconds, and the output of the band pass filter 10 can be obtained in an intervals of 1 microsecond, it equalizes data of pulses with pulse width of not less than 20 microseconds. However, in a digital calculation, since it is convenient for the digital calculation to enable division based on a bit shift operation, the adding data number is arranged as an n-th power value of 2, and in this case, it adds 32 pieces of data, that is, 25. In this case, since an output interval of the band pass filter 10 is 1 microsecond, the output interval of this equalizing operation, that is, the first time width of the first smoothing unit 12A, is 32 microseconds, which is 32 times as long as that of the band pass filter output. The outputs of this first smoothing unit 12A are inputted into the data removal equalization unit 13. The data removal equalization unit 12 divides the outputs of the first smoothing unit 13A at intervals of 32 microseconds for every specific number according to the second time width, and afterward compares sizes of data and thus removes specific data. In the case shown in FIG. 9, if it is assumed there has been a removal of a surge noise, the following operation is carried out. An arrival interval of the surge-like pulse shown in FIG. 9 is about 2 milliseconds. Thus, by dividing the data at intervals of 2 milliseconds or less and removing only pulse data acquired by sampling the surge-like pulse among them, the remainder is convertible to a power value of data without noises. That is, if it compares size relation of data in this section and eliminates two or more pieces of data which are the largest of all, the influence due to this surge-like noise is removable. Here, the minimum number of eliminating data is arranged two, because, depending on timing, there is the possibility where the surge noise with a pulse width of 20 microseconds is mixed into two output signals of the first smoothing unit 12A. Moreover, when the largest values are removed, in order to hold the average, it is also necessary to remove several pieces of the smallest values, the number of which is equal to the number of the data that had already eliminated as the largest values. In this case, two pieces of the smallest data should be removed also. Thus, the remaining data acquired by eliminating two largest values and two smallest values are equalized. In the case shown in FIG. 9, assuming that one set consists of 36 pieces of outputs of the first smoothing unit 12A, the second time width of the data removal equalization unit 13 is about 1.16 microseconds, in which at most two surge pulses are contained; therefore, by removing two larger data and two smaller data, 32 pieces of data as the remainder are equalized. In this case, the data number in one set after the removing is determined as a number of the n-th power of 2. Moreover, if an arrival interval of the surge-like pulse becomes short, it can remove the noise by, firstly, lessening the number of data added in the data removal equalization unit 13, together with evaluating a number of arriving surge-like pulses in the second time width in which the arithmetic average is calculated in the data removal equalization unit 13 and removing the double number of the evaluated arrival number of the surge-like pulse of larger values and smaller values, respectively. However, when removing data in this way, the rate of data removal may become a subject. That is, if the numbers of the output pulses of the SRNM sensor 1 are not so much, the averaged power, that is, the MSV measurement value is displayed lower than an actual average due to data removal. When the pulse number is sufficient, a certain degree of data removal is within an error range due to randomness of the data. As a result of our simulation, when the measurement minimum in the MSV measurement is set to a generating rate of the pulse, that is, 1*104 CPS, it turns out that sufficient measurement accuracy can be obtained by equalizing several percent of data in the real time. Therefore, intermittent surge noises can be removed without affecting the measurement by removing data with satisfying a necessary removal limit rate acquired from the measurement accuracy. As another function of the data removal equalization unit 13, even if the pulse generating rate of the SRNM sensor 1 is relatively low, it can obtain MSV indicated value by choosing the maximum value only. That is, if a pulse generating rate is low, there is a lot of smoothing sections in which no one piece of the sensor pulse comes. Thus, by arranging that the smoothing operation does not contain this not-coming period, the MSV indicated value can be obtained even if the pulse generating rate is low, and thus the MSV measurement minimum can be extended. This method is a measurement method using both the pulse measurement method and the MSV measurement method. However, in this method, it is necessary to compensate linearity of the MSV indicated value to the pulse generating rate by using a compensation function in which the data removing rate is used as a parameter. Next, it explains one example of this embodiment concerning an operation standard of removing data in the data removal equalization unit 13. FIG. 10 shows a simulation example of smoothing an inputted simulated neutron pulse in a period of 32 microseconds by the first smoothing unit 12A. In FIG. 10, a solid line denotes an average S after smoothing the MSV indicated value corresponds to the left vertical axis in the figure. This figure shows that the MSV indicated value after the smoothing, i.e., the second power of voltage value, changes almost in proportion to the pulse counting rate in a range of more than 1*105 CPS, which is included in the MSV measurement range. On the other hand, the MSV indicated value is not a proportionality relation to the pulse counting rate in a range between 1*104 CPS and 1*105 CPS due to influence of a circuit noise. Moreover, the dashed line shows changing of an index X denotes as:X=(S+6*σ)/S,where S is the average value of the MSV indicated value and σ is a standard deviation, corresponding to the right vertical axis in FIG. 10. The maximum of X is about 5.3 at the MSV measurement minimum, and X changes mostly between two and three in the counting rate of over 1*106 CPS by the same evaluation. On the other hand, as another evaluation method, FIG. 11 is a graph showing change of a fluctuation rate Y when a simulated neutron pulse is limited within band range between 100 kHz and 400 kHz by the band pass filter 10 and afterward equalized in 32 microseconds by the first smoothing unit 12A. The fluctuation rate Y is denoted by a formula: Y=σ/S. This figure shows that Y is not more than 0.4 in the MSV measurement range, whose minimum is 1*105 CPS, the maximum fluctuation is at the minimum of the MSV measurement range. As shown in these figures, the fluctuation of the waveform becomes large especially at around 1*105 CPS which is the minimum of the MSV measurement range. Thus, by preliminarily evaluating the indices which denotes fluctuation degree such that the above-mentioned X or Y near the pulse counting rate 1*105 where the fluctuation becomes large in a condition of the first time width of the first smoothing unit 12A, it is efficient to judge as a noise when the indices exceed the evaluated value acquired beforehand. Namely, when digital data of a signal including pulse components are equalized in 32 microseconds as the first time width by the first smoothing unit 12A, by comparing the result after the smoothing with the evaluation index and basis which are obtained by using the result of evaluating a formerly inputted signal including pulse components in the first smoothing unit 12A or a simulated neutron pulse signal beforehand, the calculated value which exceeds the evaluation basis is judged as an unusual value. For example, a threshold value which is set as eight times as large as the smoothing result of the formerly inputted pulse signal. FIG. 12 typically shows a pulse waveform containing a noise. Usual MSV value is swinging in a range surrounded with dashed lines, and the part exceeding this range can be recognized as a noise. Though this threshold value changes with the average of the MSV indicated value, evaluation using the evaluated value at the minimum of the MSV measurement at which the fluctuation is the biggest can be applied as a conservative evaluation method at everywhere in the MSV measurement range. Moreover, in the data removal equalization unit 13, there are two processing cases about the data corresponding to a part judges as a noise; one method is removing this data thoroughly, another method is replacing this data by a value in a range within the above-mentioned threshold value. In the former case, it can remove noises thoroughly but it has to evaluate the data removal rate in order to remove data within the above-mentioned permissible range of the data removal rate. In the latter case, it cannot perform perfect noise removal, but it is not necessary to evaluate the data removal rate. Hereinafter, it explains an example of the latter case for replacing the value corresponding to a noise portion in detail. FIG. 13 is a graph showing a relation of the fluctuation range of the MSV indicated value and the change width of the average value. In this figure, a solid line denotes an output signal of the MSV measurement after equalizing the pulse in 32 microseconds, a dashed line denotes a maximum change of a neutron flux, and a dot dash line denotes a change width of the fluctuation of the MSV measurement evaluated by the above-mentioned matter. The change width of the fluctuation of the MSV measurement is sufficiently larger than the change rate of the neutron flux which should be measured primarily. Therefore, it is better to suppose that a case where it exceeds the evaluated fluctuation change width in the MSV measurement is judged as an unusual value. FIG. 14(a) shows an example of change of the MSV indicated value of a pulse in which a noise exceeding this fluctuation change width, i.e., the maximum fluctuation, is induced. In this figure, an arrow A is a threshold value specified as eight times of the fluctuation change width in the MSV measurement, for example, an evaluation result after smoothing of a formerly inputted pulse. And this unusual value is replaced by a normal value, which is calculated as a product of an original maximum change rate of the neutron flux and the last sampling value. The result of this replacement is shown in FIG. 14(b). Here, it can assume that the maximum change rate of the neutron flux in a width of 32 microseconds is, for example, about 1.03, which is sufficiently smaller than the above-mentioned fluctuation range of the MSV measurement, and in this case the unusual value is permuted by a value 1.03 times as large as the last value. Therefore, by evaluating in advance a change rate in case an unusual value is detected, such as a maximum change rate of the neutron flux which should be monitored in a smoothing section, the unusual value can be removed limitedly without worsening time response of the measurement. That is, in this processing based on the fluctuation amount, by preliminarily evaluating a fluctuation range of the MSV indicated value, and by removing data exceeding this range or replacing data exceeding this range into a maximum value of the fluctuation, it becomes possible to secure an enough data number to be applied to the MSV calculation and acquire measurement result with little fluctuation. Next, the output of the data removal equalization unit 13 is inputted into the second smoothing unit 12B, and is equalized so that a fluctuation of the measured values satisfies necessary measurement accuracy and it is in a range assuring a response demand. This result of the second smoothing unit 12B is inputted into the MSV neutron evaluation unit 14 and the measured MSV value is converted to a value of a neutron flux. Moreover, in this embodiment, it is preferable to arrange a noise characteristics evaluation unit 15 for evaluating the minimum of the pulse width and arrival cycles of surge-like pulses, which are characteristics of noise waveforms, and setting the number of data used in the equalization processing in the first smoothing unit 12A, the averaging period and the number of removal data in the data removal equalization unit 13, and the time constant of smoothing filter in the second smoothing unit 12B. As mentioned above, according to this composition, even if surge-like noises are induced in the MSV measurement, by evaluating pulse duration and an arrival cycle of the surge-like noise and removing surge noise data to a certain extent satisfying the data removal rate limit permissible in the MSV measurement, the intermittent surge noises can be removed completely. Fourth Embodiment Next, a fifth embodiment of this invention is explained with reference to FIG. 15. A radiation measurement device of this embodiment shown in FIG. 15 has a CdTe sensor 16 using CdTe (cadmium, tellurium) which is a room-temperature semiconductor, as a radiation sensor. As a radiation sensor, it is also possible to use a combination of a scintillation sensor, such as NaI, and photomultiplier tubes that enable to acquire energy information, or a Ge (germanium) sensor as a semiconductor sensor. The output of the CdTe sensor 16 is inputted into a charge amplifier (CA) 17. And this charge amplifier 17 integrates electric charge of pulse components included in an input signal and converts to a pulse having a pulse height based on the amount of the electric charges to be outputted. In addition, the charge amplifier 17 supplies operating voltage to the CdTe sensor 16. An output of the charge amplifier 17 is transformed in waveform by such as a pileup rejection circuit or a pole zero cancellation circuit, which are generally used for measuring radiation energy, and afterward it is inputted to an MSV measurement unit 18, a current detector (CD) 19 and a pulse counter (PC) 20. In the MSV measurement unit 18, after restricting a frequency band, it averages the n-th powers and the average is converted to an MSV measurement value, i.e., a secondary moment value. The current detector 19 measures an average current value, which is a primary moment value, and the pulse counter 20 calculates a pulse number. The MSV measurement value, the current measurement value, and the pulse enumerated number are inputted into an energy evaluation unit 21, respectively, and the energy evaluation unit 21 evaluates average radiation energy based on a ratio of the MSV value to the number of pulses or a ratio of MSV value to the direct current value, that is, a ratio of the secondary moment and the primary moment. This average energy value and the above-mentioned measured values are inputted into a dosage evaluation unit 22, and thus they are converted to an irradiation dose, or an absorbed dose in a substance, or a dose equivalent including a risk rate to a human body. The output of the charge amplifier 17 is a pulse having a peak value proportional to a radiation energy absorbed in the CdTe sensor 16. Therefore, suppose that a probability where the reaction occurs is N and the absorption energy is q, the MSV value, the pulse enumerated number and the current value can be approximated by the following formulas: MSV value: k1*q2*N, n-th moment value: kn*qn*N, Pulse enumerated number: k2*N, and Direct current value (primary moment value): k0*q*N, Here, k0, k1, k2 and kn are compensation coefficients, respectively. And their ratios are: MSV value/pulse enumerated number=k1*q2 (generally, kn*qn), MSV value/direct-current value=(k1/k0)*4, and n-th moment value/n′ moment value=kn*qn−n′/kn. Therefore, it can presume the absorption energy in a crystal by evaluating these compensation coefficients k0, k1, k2 and kn, etc., beforehand and using ratios of these measured values. FIG. 16 is a plotted graph showing a relation of the pulse enumerated number and the MSV value (shown in a vertical axis) and a dosage (shown in a horizontal axis) measured by a commercial radiation survey meter when it measures radiations of various radioactive elements, that is, radiations having different energy respectively, by the CdTe sensor 16. Generally, the surveymeter, etc., is adjusted in internal compensation coefficients or shielded, so that sensitivity characteristics to the radiation energy agree with an evaluation curve of the dose equivalent to the radiation energy. That is, although a pulse counting rate becomes large to a radiation having low energy since its absorption energy when one radiation is irradiated is low, the dosage of one radiation in this case becomes low. On the contrary, although a pulse counting rate becomes low to a radiation having high energy, the dosage becomes large since an amount of electric charges generated by one radiation is large. Thus, it is adjusted by shielding, etc., so that the pulse counting number or the current value sensitivity becomes the same as a contribution rate to the dose equivalent. Since the case shown in FIG. 16 omits this sensitivity compensation, the pulse enumerated number is large in a radiation with low energy, and the pulse enumerated number and the MSV measurement value become random to the dose equivalent. However, if it is plotted as characteristics to the dosage of a ratio of the MSV value to the pulse enumerated number, as shown in FIG. 11, it becomes monotonous characteristics to the dose equivalent. Thus, it is possible to convert the ratio of these to the dose equivalent by evaluating these characteristics in advance. Similarly, since the ratio of the MSV value to the pulse enumerated number serves also as monotonous characteristics to incident energy, it is possible to presume average incident radiation energy by evaluating these characteristics in advance. In this case, it becomes possible to evaluate an absorbed dosage at each part of a human body to the radiation energy more accurately by using absorption characteristics of the part of a human body. Furthermore, there are two cases to evaluate the dosage by the pulse measurement. One case is a method for converting energy information of the incident radiation acquired by measuring a pulse height distribution of the pulse to the dosage, and another case is a method for equalizing the sensitivity of the pulse measurement and dosage response characteristics by devising structure itself of the above-mentioned sensor. Moreover, as a way of evaluating the dosage by the current value, there is a method of adjusting a sensitivity response by the devising the sensor structure mentioned as the latter method of the above-mentioned cases. Therefore, it can perform still more accurate dosage evaluation by using both these common techniques and the dosage evaluation method in this embodiment. That is, for example, if it uses the evaluated compensation function in this embodiment after adjusting the sensitivity characteristics of the sensor to some extent independently, it becomes possible to perform still more exact dosage evaluation. Furthermore, if the pulse is piled up in a high counting rate with dropout count, it cannot evaluate an accurate dosage by the method of converting the acquired pulse height information into the dosage mentioned as the former method of the above-mentioned cases. And in the latter shielding method of the above-mentioned cases, it must rectify a number of the dropout count. However, in this embodiment, it is possible to evaluate the dosage in the MSV measurement and measure in a large range by performing the pulse measurement and the MSV measurement simultaneously, even when the pulse measurement is saturated by the pileup. Although it needs to rectify the pileup effect of the pulse measurement in the presumption of the average energy in this case, the error can be suppressed in a range which can be neglected by making the sensitivity of the sensor itself approximate to the dosage response to some extent. Moreover, if it uses a ratio of the current value to the pulse enumerated number or a ratio of the MSV value to the current value, as well as the ratio of the MSV value to the pulse enumerated value, it can presume the average radiation energy by acquiring a compensation function similar to that of the case mentioned above. In this way, according to this embodiment, by using both the n-th moment value and the pulse measurement together, it can evaluate the dosage accurately based on the presumption of the average incident energy. Moreover, even if it is in a condition occurring counting error due to the pileup of the pulses, by using the ratio of the n-th moment to another n-th moment, it can presume the average energy similarly, and the dosage evaluation is carried out exactly in a measurement range larger than that of the conventional method. Furthermore, hereinafter it explains a deformed example of this embodiment. Here it can presume energy distribution by calculating the first power value, that is an average current, and the second value, the third value, etc., and the n-th power value, instead of the MSV value, and calculating a compensation function of each value, respectively, and solving a reverse matrix of each compensation function, respectively. That is, the measurement value of each n-th moment can be expressed as follows:x1=a1[1:n]*E[n:1], (equivalent to the current measurement value)x2=a2[1:n]*E[n:1], (equivalent to the MSV measurement value)x3=a3[1:n]*E[n:1], . . . ,xn=an[1:n]*E[n:1], wherein, xk: k-th moment value [scalar quantity], ak: Response matrix [matrix with one line and n columns], E: Energy distribution [matrix with n lines and one column]. Here, it can denote the relation of the matrices X and E by using a matrix A with n lines and n columns, as follows:X[n:1]=A[n:n]*E[n:1]. Thus, the radiation energy distribution can be acquired by solving a reverse matrix of the matrix A, such as:E[n:1]=A−1[n:n]*X[n:1]. However, in this moment measurement from the first power to the n-th power, it is sufficient to select the number of the moments corresponding to a necessary energy bandwidth from the above-mentioned formulas, and it can consist only of alternating current measurement means by removing the average current value as the first moment among the moment values. As mentioned above, by combining the MSV measurement and one of the pulse measurement and the current measurement, it can presume the average radiation energy by the ratio and convert it to the dosage. This can easily realize characteristics which are more similar to the dosage response by using together with a conventional technique of rectifying the sensitivity by changing the sensor structure. Moreover, it is not necessary to sort out the pulse height for realizing with easy composition, compared with the conventional technique of computed the dosage by questing the pulse height. Furthermore, it can reconstruct the radiation energy by using two or more n-th moment values, and it can measure the radiation energy distribution even in the case of high counting rate making the pulse measurement difficult, and evaluate the dosage more accurately from this information. In this way, by using this method independently or combined with the conventional dosage evaluation method, it can provide a radiation measurement device for collectively monitoring in a wider range more exactly. As explained above, according to the radiation measurement device of the above-mentioned embodiments in this invention, it can reduce a bipolar circuit noise with a small signal level and an alpha ray noise of the sensor and a ratio of these noises to the sensor pulse, by calculating the n-th power values of the pulse waveform and discriminating with the values, and thus it can measure a sensor signal which is mixed in the circuit noise in the conventional method. The foregoing discussion discloses and describes merely a number of exemplary embodiments of the present invention. As will be understood by those skilled in the art, the present invention may be embodied in other specific forms without departing from the spirit or essential characteristics thereof. Accordingly, the disclosure of the present invention is intended to be illustrative of, but not limiting to, the scope of the invention, which is set forth in the following claims. Thus, the present invention may be embodied in various ways within the scope of the spirit of the invention. |
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claims | 1. An X-ray optical apparatus, comprising:an X-ray reflective structure in which at least three reflective substrates are laminated so as to match both edges with an interval and an X-ray which is incident into an X-ray passage formed by a space, both sides of the passage being put between the reflective substrates, is reflected from the reflective substrate at both sides of the X-ray passage and emitted from the X-ray passage,wherein the at least three reflective substrates are arranged to have a constant and equal thickness, andwherein when an edge of the X-ray reflective structure is an inlet of the X-ray and the other edge is an outlet of the X-ray, a pitch of the reflective substrates at the outlet is larger than a pitch at the inlet, andwherein spacers arranged to have different heights are disposed between the reflective substrates, so that the pitch of the reflective substrates at the outlet side is larger than the pitch at the inlet side. 2. The X-ray optical apparatus according to claim 1, wherein the spacers are arranged to have a pillar shape and are disposed between the reflective substrates with a predetermined interval. 3. The X-ray optical apparatus according to claim 2, wherein the spacers are disposed at the same position on different layers. 4. The X-ray optical apparatus according to claim 1, wherein the reflective substrates and the spacers are integrally formed by etching a glass substrate. 5. The X-ray optical apparatus according to claim 1, further comprising:an X-ray source,wherein if a virtual plane is set in a position which is separated from the reflective substrates at both sides of the X-ray passage with the same distance, the X-ray source is disposed on tangential planes of a plurality of virtual planes at the inlet and tangential planes of the plurality of virtual planes at the outlet are approximately parallel. 6. The X-ray optical apparatus according to claim 1, further comprising:an X-ray source,wherein if a virtual plane is set in a position which is separated from the reflective substrates at both sides of the X-ray passage with the same distance, the X-ray source is disposed on tangential planes of a plurality of virtual planes at the inlet and tangential planes of the plurality of virtual planes at the outlet intersect on a common straight line. |
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abstract | Apparatus and methods for shipping hazardous materials, such as radioactive liquids and materials, that comply with certain international and national regulations for the transport of hazardous materials. According to one embodiment, hazardous material samples are placed inside inner containers, such as glass bottles. Inner containers are placed inside secondary containers, surrounded and cushioned by high density polyethylene (“HDPE”) foam inserts. Secondary containers are placed inside cutouts in a HDPE custom insert. Custom insert is inside an outer container made with HDPE and held together with extruded aluminum hardware, closed-end aluminum and steel rivets, and a sealant applied to the seams. |
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abstract | The invention provides a method for collimating a radiation beam, the method comprising subjecting the beam to a collimator that yaws and pitches, either separately or simultaneously relative to the incident angle of the beam. Also provided is a system for collimating radiation beams, the system comprising a collimator body, and a stage for pitching and yawing the body. A feature of the invention is that a single, compact mask body defines one or a plurality of collimators having no moving surfaces relative to each other, whereby the entire mask body is moved about a point in space to provide various collimator opening dimensions to oncoming radiation beams. |
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046719199 | description | DESCRIPTION OF THE PREFERRED EMBODIMENTS An embodiment of the present invention is illustrated in FIG. 2 and includes neutron sensors 10 and current-to-voltage amplifiers 15 similar to those used in the prior art circuit illustrated in Fig. 1. According to the present invention, both of the neutron sensors 10 may be connected to a single one of the current-to-voltage amplifiers 15, as indicated by the dashed line and as in the prior art circuit of FIG. 1, or each neutron sensor 10 may supply a current signal as the sole input to one of the current-to-voltage amplifiers 15. The analog voltage output by the current-to-voltage amplifiers 15 is supplied to an analog/digital converter 40 for input to a microprocessor 45. The analog/digital converter 40 and microprocessor 45 may be implemented on a single board computer such as an Intel 88/40. Depending on the particular analog/digital converter 40 and microprocessor 45 and monitoring system requirements, separate current-to-voltage amplifiers 15 may each be connected to separate analog-digital converters 40 or, as indicated by the dashed line, be connected to the same analog-digital converter on separate channels. Similarly, there may be one or more microprocessors 45 for each analog/digital converter 40. Mere replacement of the analog circuits illustrated in FIG. 1 by the digital circuitry illustrated in FIG. 2, may simplify the alignment process and the speed with which the power level monitoring circuit responds to transients; however, there is no automatic reduction in noise. It is possible to reduce the effects of nuclear noise by proper selection of the algorithm performed by the microprocessor 45. The selection of an algorithm is governed by several factors. First, the algorithm should be capable of reducing noise without total loss of the ability to detect transients. Secondly, power level monitoring circuits which operate in the power range of a nuclear reactor are required to supply a signal indicating rate of change of the power level, so that a "trip" can be generated when the rate of change has a magnitude above a specified value, i.e., indicating a sudden change in power, which may be a transient condition. One algorithm which meets these requirements is defined by alpha-beta tracker equations which are commonly used in radar applications. Application of the alpha-beta tracker equations to processing in radar systems is described in J. A. Cadzow, Discrete-Time Systems, Prentiss-Hall, 1973, sections 2.6 (pages 63-66) and 8.11 (pages 272-278). Applying the equations described in Cadzow to the power level monitoring circuit illustrated in FIG. 2, the digital voltage output by the analog/digital converter 40 can be represented by f(k). The power level p(k), rate of change of power level p(k) and predicted power level p.sub.p (k) are defined by equations (1)-(3) below. EQU p.sub.p =p(k-1)+T p(k-1) (1) EQU p(k)=p.sub.p (k)-.alpha.[f(k)-p.sub.p (k)] (2) ##EQU1## The power level p(k) is an estimate or smoothed output for the current sampling period in which the effects of noise have been reduced. The predicted power level p.sub.p (k) is a prediction of the estimated power p(k) for the immediately following sampling period. The length of the sampling period is represented by T, .alpha. and .beta. are constants which determine the dynamic response of the power level monitor. The interrelationship of equations (1)-(3) is visually represented by the block diagram illustrated in FIG. 3. The input sample signal f(k) is converted by adder 110 and multiplied by constants .alpha. and .beta./T in multipliers 120 and 130. The resulting signals are input to adders 140 and 150, respectively. The outputs of adders 140 and 150 are supplied to registers 160, 170 and 180 which provide a delay of T. The output of register 160 is summed with the output of multiplier 130 to provide the rate of change of the power level p(k). The output of register 180 is multiplied by the length of the sampling period T in multiplier 190 prior to being summed with the output of register 170 in adder 200 to provide the predicted power level p.sub.p (k). The predicted power level p.sub.p (k) is multiplied by negative one so that it is subtracted from the sampled signal f(k) by adder 110 and is also summed with the output of multiplier 120 to produce the smoothed power level p(k). Selection of appropriate values for the constants .alpha. and .beta. is explained in Cadzow in section 8.11 (pages 272-278) using the Z-transform which is 8.11 (pages 272-278) using the Z-transform which is throughly discussed on pages 144-175 of Cadzow. The Z-transform of equations (1)-(3) are illustrated as a block diagram in FIG. 3B and appear below as equations (4)-(6). EQU P.sub.p (z)=z.sup.-1 p(z)+z.sup.-1 T P(z) (4) EQU P(z)=P.sub.p (z)+.alpha.[F(z)-P.sub.p (z)] (5) ##EQU2## Since the outputs illustrated in FIG. 3B are all derived from a single input, the following transfer functions H.sub.1 (z)-H.sub.3 (z) can be defined. EQU P(z)=H.sub.1 (z)F(z) (7) EQU P(z)=H.sub.2 (z)F(z) (8) EQU P.sub.p (z)=H.sub.3 (z)F(z) (9) Dividing both sides of equations (4)-(6) by the Z-transform F(z) of the input signal f(k), incorporating the transfer function relationships of equations (7)-(9) and rearranging the terms, results in the following equations (10)-(12). EQU z.sup.-1 H.sub.1 (z)+z.sup.-1 T H.sub.2 (z)-H.sub.3 (z)=0 (10) EQU H.sub.1 (z)-(1-.alpha.)H.sub.3 (z)=.alpha. (11) ##EQU3## Solving equations (10)-(12) for H.sub.1 (z)-H.sub.3 (z) results in the following equations (13)-(15). ##EQU4## The denominator polynomial which is common to all three of the fractions above is known as the characteristic equation which defines the system poles. Solving for the poles of the characteristic equation yields equation (16) below. ##EQU5## Assuming a critically damped system is desired, the term (.beta..sup.2 +.alpha..sup.2 +2.alpha..beta.-4.beta.) is set to zero with the result that .alpha.=2.sqroot..beta.-.beta.. Substituting for in equations (13)-(15) produces the following equations (17)-(19). ##EQU6## Thus, a critically damped power monitor using alpha-beta tracker equations has a double pole of z=1-.sqroot..beta.. Implementation of an alpha-beta tracker power level monitor requires selection of a sampling period length T and a value for .beta., from which the value of .alpha. can be found. The sampling period length T will be determined by the speed of the neutron sensor 10, analog/digital converter 40, and the requirements of the equipment which receives the signals output by the microprocessor. A discussion of how to select the value of .beta. can be found on page 278 of Cadzow and in Benedict, T. R. and Bordner, G. W., "Synthesis of an Optimal Set of Radar Track--While Scan Smoothing Equations," IRE Transactions on Automatic Control, Vol. AC-7, No. 4 (July, 1962) pages 27-32. The value of .beta. affects the degree of noise reduction and system response speed. For applications such as data logging of the neutron flux in a nuclear reactor, a value of .beta. equal to or very close to zero is preferable, because the effects of noise will be minimized. However, the response time will be very slow. Therefore, in neutron flux monitors which must generate trip signals, the value of .beta. is increased (up to a maximum of 1) until statisfactory system response time is achieved. The amount of noise suppression provided by an alpha-beta tracker is reduced as .beta. is increased; therefore, .beta. should be selected to be as small as possible while meeting the system response time requirements. Once the value of has been selected, the value of .alpha. can be found as 2.sqroot..beta.-.beta. and a program for microprocessor 45 can be easily written from the block diagram in FIG. 3A. When implemented, a power level monitor using alpha-beta tracker equations will provide noise suppression and "fast follow" capibility for responding to transients in the neutron flux. In addition, both the rate of power level change p(k) and a predicted next power level p.sub.p (k) are automatically produced by the alpha-beta tracker equations. Also, alignment of such a power level monitor is considerably simplified due to the noise suppression capabilities of the alpha-beta tracker equations and the use of digital processing which eliminates the need for adjusting a variable resistor in a rate/lag circuit 20 as in the prior art. The many features and advantages of the present invention are apparent from the detailed specification, and thus it is intended by the appended claims to cover all such features and advantages of the power level monitor which fall within the true spirit and scope of the invention. Further, since numerous modifications and changes will readily occur to those skilled in the art, it is not desired to limit the invention to the exact construction and operation illustrated and described, accordingly, all suitable modifications and equivalents may be resorted to, falling within the scope of the invention. |
047042451 | claims | 1. A method for monitoring a break in an ion adsorption apparatus in a core water purification system of a nuclear reactor with a primary water coolant containing sodium ions by detecting a break point of the ion adsorption apparatus using ion exchange resin, thereby determining a time for regenerating or exchanging the resin, which comprises making a sampling ion species having a weaker selective adsorbability to the ion exchange resin than that of a target ion species to be adsorbed also present in water to be treated, and detecting leakage of the sampling ion species at the downstream side of the adsorption apparatus, thereby determining the break point of the ion exchange resin; said sampling ion species being radioactive sodium ions and said target ions species being radioactive cobalt ions. 2. A method according to claim 3, wherein a change in concentration of the sampling ion species between the upstream side and the downstream side of the adsorption apparatus, thereby determining a time (Ts).sub.1 at which the outlet concentration of the sampling ion species starts to increase from an initial constant value and a time (Ts).sub.2 at which a ratio of outlet concentration to inlet concentration of the sampling ion species starts to exceed 1, and determining a break time T.sub.B of the target ion species to be adsorbed according to the following formula: EQU T.sub.B =(Ts).sub.2 (Ko/Ks)-{(Ts).sub.2 -(Ts).sub.1 }(Ks/Ko) 3. A method according to claim 1, wherein concentrations of the sampling ion species are detected at the upstream side and the downstream side of the adsorption apparatus, and the break time is a period from a time at which the ratio of outlet concentration to inlet concentration starts to exceed 1 to a time at which the ratio returns to 1. 4. An apparatus for monitoring a break of ion adsorption apparatus in a core water purification system of a nuclear reactor with a primary water coolant containing sodium ions, which comprises an adsorption unit with ion exchange resin, two radioactive detectors, one at the upstream side and the other at the downstream side of the adsorption unit, and a unit for calculating a ratio of concentrations detected by the two detectors; means for making a sampling ion species having a weaker selective adsorbability to the ion exchange resin than that of a target ion species to be adsorbed also present in water to be treated, said sampling ion species being radioactive sodium ions and said target ion species being radioactive cobalt ions; and means for detecting leakage of he sampling ion species at the downstream side of the adsorption apparatus including said two radioactive detectors and the unit for calculating a ratio of concentrations, the concentrations of the sampling ion species being detected at the upstream side and the downstream side of the adsorption apparatus by said detectors and the break time being a period from a time at which the ratio of outlet concentration to inlet concentration of the sampling ion species starts to exceed 1 to a time at which the ratio returns to 1. 5. A method according to claim 3, wherein the concentrations of the radioactive sodium ions are determined by .gamma.-ray scanning of piping containing the water to be treated upstream and the water treated downstream of the adsorption apparatus. |
claims | 1. Device for measuring the drop time of a control rod into the core of a nuclear reactor comprising: a vessel closed in its upper part by a closure head and containing the reactor core including fuel assemblies positioned in vertical adjacent arrangements and having vertical guide tubes for receiving neutron absorbers of a plurality of control rods, each control rod including a neutron absorber rod cluster; a drive shaft connected at one of its ends to the control rod cluster; a plurality of tubular enclosures placed vertically above the closure head of the vessel, each tubular enclosure having a mechanism for moving a control rod by means of its vertical drive shaft, in the axial vertical direction of the tubular enclosure and for releasing the drive shaft in order to ensure that it drops into the position of maximum insertion in the nuclear reactor core; at least one electrical measurement winding surrounding each tubular enclosure, substantially over its entire axial length and connected to a measurement cable for picking up a measurement signal at the terminals of the winding; the control rods being grouped together in subassemblies and each electrical winding for each of the rods of a subassembly being connected via a measurement cable to a cabinet of rod position instrumentation containing a plurality of control rod position instrumentation sensors; a real-time measurement-signal acquisition module; and means for processing the measurement signals being connected to the output of the acquisition module for measuring the drop time of the nuclear reactor control rods; wherein each cabinet of the rod position instrumentation includes, for each measurement cable connected at its input to the winding of a control rod, an interface box connected permanently to an output of the measurement cable and to an input of a transmission cable for continuously transmitting a measurement signal to the real time acquisition module without interruption of supply power to the winding. 2. Device according to claim 1 , wherein each interface box comprises an attenuator for attenuating the level of the measurement signal transmitted to the real-time acquisition module. claim 1 3. Device according to claim 1 , wherein the measurement signal is filtered to overcome the effect on the drop time measurement of a supply voltage of the winding provided through a cabinet of the rod position instrumentation. claim 1 4. Device according to claim 1 , wherein each of the transmission cables is connected to the real-time acquisition module via an isolation box. claim 1 5. Device according to claim 1 , comprising four cabinets containing nuclear-reactor control-rod position instrumentation sensors, each rod being associated with one part of the nuclear reactor core having a quadrant-shaped cross section, the nuclear reactor core including four quadrants defined by two mutually perpendicular planes of symmetry passing through the vertical axis of the nuclear reactor core, the electrical measurement windings for each of the control rods in one part of the nuclear reactor core being connected to one of the four cabinets containing control-rod position instrumentation sensors and the interface boxes connected to the measurement cables of the electrical windings for the control rods being placed in the cabinets containing the rod position instrumentation sensors. claim 1 6. Device according to claim 5 , comprising four DC isolation boxes, the input of each being connected to an output of a respective interface box, thereby transferring the measurement signals from the control rods to the signal acquisition module. claim 5 7. Device according to claim 1 , wherein each interface box is placed inside a nuclear reactor protection room and the measurement-signal acquisition module and the means for processing the signals are placed in a room from which the movement of the control rods is controlled. claim 1 8. Device according to claim 1 , wherein the means for processing voltage signals, which are connected to the acquisition module, comprise a microcomputer for calculating the drop time of the control rods. claim 1 9. Device according to claim 1 , wherein the electrical measurement windings comprise primary windings of transformers that measure the position of the control rod. claim 1 |
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description | The Nuclear Regulatory Commission (NRC) has established design criteria as part of the cod of federal regulations (i.e. 10CFR50 appendix A). Design criterion #26 states: “Two independent reactivity control systems of different design principles shall be provided. One of the systems shall use control rods, preferably including a positive means for inserting the rods, and shall be capable of reliably controlling reactivity changes to assure that under conditions of normal operation, including anticipated operational occurrences, and with appropriate margin for malfunctions such as stuck rods, specified acceptable fuel design limits are not exceeded. The second reactivity control system shall be capable of reliably controlling the rate of reactivity changes resulting from planned, normal power changes (including xenon burnout) to assure acceptable fuel design limits are not exceeded. One of the systems shall be capable of holding the reactor core subcritical under cold conditions.” Previous designs all complied with the requirement to provide a rod control system. Boiling Water Reactor (BWR) used a separate reactor core flow control which effectively manipulated the moderator (i.e. coolant properties) through the core. Pressurized Water Reactor (PWR) designs used a mixture of boron within the primary coolant which also effected how the moderator could control the nuclear chain reactions within the core. Beyond these two types of manipulation of the water within the core of a reactor, there was no other way to sufficiently provide separate control through all the range of power levels expected by the NRC. The only other moderator control systems are applicable to light water reactors by changing the properties of the water within the core (i.e. phase changes in a BWR and chemical addition in a PWR). Void content changes in the water of a BWR form a very rapid feedback loop for power ascension and reduction. The use of poison control for a PWR is a much slower process. This invention changes the physical geometry around the core and therefore can always be relied upon to perform its function. This reactor control system can be applied to all types of reactors regardless of design. This design for reactor control is used in addition to the necessary rod control system in accordance with the 10CFR50 design criteria. This allows for the requirements of the NRC to be met along with the ability for dual control on power control of any type reactor regardless of process output from the secondary plants. 10: is the carriage brace; 12: are the suspension rods; 14: is the vertical gearing rod; 16: is the vertical gearing rod motor; 18: is the center gearing; 20: is the carriage brace motor; 22: is the neutron detection system input (incore or excore detectors); 24: is the control logic for movement; 26: are the neutron reflector/moderator blocks; 28: is the suspension rod brace (to control downward movement of RM blocks; 30: is the release locks; 32: is the carriage; 34: is the reactor core; 36: is the reactor containment; and 38: is the reactor. As stated above, The Nuclear Regulatory Commission (NRC) has established design criteria as part of the cod of federal regulations (i.e. 10CFR50 appendix A). Design criterion #26 states: “Two independent reactivity control systems of different design principles shall be provided. One of the systems shall use control rods, preferably including a positive means for inserting the rods, and shall be capable of reliably controlling reactivity changes to assure that under conditions of normal operation, including anticipated operational occurrences, and with appropriate margin for malfunctions such as stuck rods, specified acceptable fuel design limits are not exceeded. The second reactivity control system shall be capable of reliably controlling the rate of reactivity changes resulting from planned, normal power changes (including xenon burnout) to assure acceptable fuel design limits are not exceeded. One of the systems shall be capable of holding the reactor core subcritical under cold conditions.” Previous designs all complied with the requirement to provide a rod control system. Boiling Water Reactor (BWR) used a separate reactor core flow control which effectively manipulated the moderator (i.e. coolant properties) through the core. Pressurized Water Reactor (PWR) designs used a mixture of boron within the primary coolant which also effected how the moderator could control the nuclear chain reactions within the core. Beyond these two types of manipulation of the water within the core of a reactor, there was no other way to sufficiently provide separate control through all the range of power levels expected by the NRC. The invention claimed here solves this problem. This invention changes the core geometry in order to positively control the nuclear reaction and is potentially applicable to all known types of reactors, even light water reactors (LWR). The ability to provide a control system based on the movable geometry of the reactor allows for a separate and distinct reactor control system. This control system provides an ability to change the core geometry to increase or decrease, as necessary, the neutron coupling within the core. In the case of light water reactors, this would be a moderating and reflecting property. In the case of fast neutron reactors, this would be mostly the reflecting properties to maintain the core neutron inventory. The claimed invention differs from what currently exists. This adds to the safety of the reactor. Thus creating a safer reactor by orders of magnitude higher than what exists today. The core can be shut down by control rods (neutron absorption) in the core and also by removing the necessary core geometry for moderation and reflection of those neutrons into the core in order to be able to sustain the nuclear reaction through this invention. This limitation of core geometry also helps limit the consequences of a rod withdrawal accident thus also increasing reactor safety. This invention is an improvement on what currently exists. This adds to the safety of the reactor. Thus creating a safer reactor by orders of magnitude higher than what exists today. The core can be shut down by control rods (neutron absorption) in the core and also by removing the necessary core geometry for moderation and reflection of those neutrons into the core in order to be able to sustain the nuclear reaction through this invention. This limitation of core geometry also helps limit the consequences of a rod withdrawal accident thus also increasing reactor safety. The previous method of changing the properties of the moderator can only be applied to very specific light water reactor designs. Other types of reactors such as a high temperature gas cooled reactor (HTGR), Liquid sodium fast breeder reactor (FBR), or liquid fluoride thorium reactor (LFTR) cannot use water as cooling and therefore cannot use either previous method. Liquid salt and liquid metal type reactors also do not use water as a moderator/reflector and therefore cannot use any systems that change the properties of water. This reactor control system can be applied to all types of reactors regardless of design. This design for reactor control is used in addition to the necessary rod control system in accordance with the 10CFR50 design criteria. This allows for the requirements of the NRC to be met along with the ability for dual control on power control of any type reactor regardless of process output from the secondary plants. The Version of the Invention Discussed Here Includes: A. Carriage Brace B. Suspension rods C. Vertical gearing rod D. Vertical gearing rod motor (provides vertical movement) E. Center gearing F. Carriage brace motor (provides horizontal movement) G. Neutron detection system input H. Control logic for movement I. Neutron reflector/moderator blocks J. Suspension rod brace (downward mechanical stops) K. Release locks'Relationship Between the Components: The RM blocks are suspended from the carriage brace by the suspension rods. The carriage brace is connected together (two separate halves) through the center gearing. The RM blocks are positioned with the vertical gearing rod through the center of the block and connected to the suspension carriage. The center gearing provides a stationary center point of the whole structure. The carriage brace motor is attached to the gear teeth of the carriage brace. Suspension rod braces are attached to the rods underneath the RM blocks. Release locks are located on the vertical gearing rods at the top of the RM blocks. These are electrically controlled by the control logic system associated with the RM control. Neutron system input is delivered into the RM control system to provide trip set points. How the Invention Works: The RM blocks from the geometry for the core to be neutronically coupled for the nuclear chain reaction to work. The RM blocks are suspended on the suspension rods at core height. These provide vertical and horizontal stability to the blocks. The vertical gearing rods are in the center of the blocks and allow for vertical positioning of the blocks. The carriage braces are connected together by center gears much like an expandable kitchen table arrangement. The opposite sides of the brace are arranged next to the center gear. When one side of the brace is moved, this turns the fixed center gear which then moves the opposite side of the carriage brace. Since the RM blocks are suspended from this carriage, the blocks can move towards or away from the reactor core. When the appropriate distance is reached, the core is then geometrically configured to allow the neutron chain reaction to be sustained. The core power detectors are connected to the control system for the purpose of control. There is also a trip set point that can be reached which could be coupled to the existing high power/oscillating power set points. The result of actuating one of these power tripping points would activate the latch release mechanisms of the RM blocks. The blocks would slide down and out of the way. There would also be a separate mechanism that would withdraw the RM carriage to the most open position (i.e. the RM block are moved outward away from the reactor vessel as far as carriage allows. The RM blocks are stopped by the suspension rod braces that do not allow the RM blocks to fall the whole way to the bottom of the core. Were the core to ever reach a molten state (in the case of a LWR) or already be in a molten state (such as a LFTR), then the corium would find its way to the bottom of the reactor (or separate holding tank in the case of the LFTR) well past the bottom of the RM blocks thus ensuring that the function of providing neutron coupling is not still supplied to the core in that condition. This principle works well with a pool type liquid sodium reactor as well. Since the carriage and all electrical components sit up at the carriage which would be suspended over the pool, the RM blocks could be moved back and forth within the pool with ease. Since they are already suspended on movable carriages with the ability to pull them up and down, it would be easy to pull them to the surface of the pool and then remotely attach a lifting device to them for removal from the pool. The general control of the blocks will be through manual controllers. The trip logic controls the block drop and withdrawal function for the system. IF the input from the neutron instrumentation reaches the trip set point, THEN the logic will unlatch the RM blocks and allow them to fall to the lowest position on the cable while simultaneously backing the carriages away from the core. The upward movement of the RM blocks is controlled manually. How to Make the Invention: The invention is made and assembled in accordance with the description above. The RM blocks are made of the most appropriate reflective material with moderation properties. This is usually a form of carbon block. Suspended tanks of water or concrete can prove difficult to manage from a material property and weight perspective. The blocks are suspended from the carriage after the carriage is installed. Control system logic is connected to the motors that control the upward/downward motion of the RM block and also the inward/outward motion of the control carriage. Power instrumentation for the reactor is than input into the control system. Logic is constructed so that the protective actions are actuated should the set points be reached. The system components are necessary. Elements that could be added might include a neutron absorption shield that might fall in front of the RM blocks to even more quickly absorb the core neutrons. This function is normally considered the function of the control rod system's function. Duplication of this function might add a marginal amount of safety but could turn out to be cost prohibitive in relation to the rod control system. The ability to remove the core's neutron coupling provides a completely separate and distinct method for control of reactor power. The blocks could be made into petals like a flower and radially drop back out of the way to remove them from service. The RM blocks could be shaped in a configuration like the iris of a camera and then rotated out of the way much like opening that iris in order to remove their coupling function. In the case of pool type reactors, the use of RM rods or correctly configured rods could be used interspersed with the poison control rods. How to Use the Invention: The core needs to be controlled through the establishment of the correct geometry and the removal of the control rods in order to allow the reactors chain reaction to occur. This could be done with the establishment of the correct geometry first and then the withdrawal of the control rods. Once the section of the core that has the RM blocks surrounding it is used up the blocks are moved vertically upward (in the case of the movable blocks) or the core is finished just as conventional cores are today. Since this is an alternate method of control, it is possible to establish a certain rod line (i.e. control rods are withdrawn to a certain level) and then the RM blocks are moved into position horizontally toward the core to provide coupling. As more coupling is desired, the blocks can move even closer to the outer wall of the reactor vessel (or in the case of the liquid pool types, simply closer to the fuel arrangement). By using RM blocks that only convert a certain segment of the core to a usable geometry, the severity of a rod withdrawal accident is greatly reduced. The portion of the core that could produce power is limited to the length of the RM block. This design could significantly reduce the impact or any rod withdrawal accident while also providing a methodology of removing the blocks from the affected part of the core in order to contain this accident by an even greater degree. The reactor is monitored and controlled with the same overall set points for safety as before, however, now there are two distinct systems that can control reactor power and can instantaneously shutdown the reactor. The same power level instrumentation is used as an input into the rod control system as for this new reactor control system. This ensures that there is no possibility for offset values to trip the rod control system or the new reactor control system inadvertently. The trip system relies on gravity to perform most of the function of removing the RM blocks from the power production area of the core. Most of the maintenance and repair of this system can be done without ever opening the reactor vessel. This ensures that operational problems are more easily accomplished. This leads to a higher reliability for this system which also helps to lower the overall risks of operation. This type of system allows for more fuel to be loaded in the reactor core with only a certain segment being covered by the length of the RM blocks. As the core ages and the bottom fuel is used, the RM blocks could be raised through the vertical positioning rods and thereby activating fuel higher up with the fuel rod assemblies. This would allow the core to operate for a longer period between refueling outages. By using this positioning method outages could be theoretically changed from 1.5-2 years out to 4-5 years. Thus this could significantly increase the revenue generation of the plant. In the cases with liquid type fuels or ball type fuels that can be refueled through a feed-and-bleed method, this would have no effect whatsoever. |
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description | This application is a Divisional of pending U.S. application Ser. No. 09/932,531, filed on Aug. 17, 2001. This invention was made with United States Government support under Contract No. DE-AC07-99ID13727 awarded by the United States Department of Energy. The United States Government has certain rights in the invention. This invention relates generally to non-destructive testing of materials and more specifically to methods and apparatus for performing non-destructive testing of materials using positron annihilation. Non-destructive testing is the name commonly used to identify any of a wide variety of techniques that may be utilized to examine materials for defects without requiring that the materials first be destroyed. Such non-destructive testing of materials is advantageous in that all materials or products may be tested for defects. That is, after testing, acceptable (e.g., substantially defect-free) materials may be placed in service, while the defective materials may be re-worked or scrapped, as may be required. Non-destructive testing techniques are also advantageous in that materials already in service may be tested or examined in-situ, thereby allowing for the early identification of materials or components that may be subject to in-service failure. The ability to test or examine new or in-service materials has made non-destructive testing techniques of extreme importance in safety or failure sensitive technologies, such as, for example, in aviation and space technologies, as well as in nuclear power generation systems. One type of non-destructive testing technique, generally referred to as positron annihilation, is particularly promising in that it is theoretically capable of detecting fatigue damage in metals at its earliest stages. While many different positron annihilation techniques exist, as will be described below, all involve the detection of positron annihilation events in order to ascertain certain information about the material or object being tested. In one type of positron annihilation technique, positrons from a radioactive source (e.g., 22 Na, 68 Ge, or 58 Co) are directed towards the material to be tested. Upon reaching the material, the positrons are rapidly “thermalized.” That is, the positrons rapidly lose most of their kinetic energy by collisions with ions and free electrons present at or near the surface of the material. After being thermalized, the positrons then annihilate with electrons in the material. During the diffusion process, the positrons are repelled by positively charged nuclei, and thus tend to migrate toward defects such as dislocations in the lattice sites where the distance to positively charged nuclei is greater. In principle, positrons may be trapped at any type of lattice defect having an attractive electronic potential. Most such lattice defects are socalled “open volume” defects and include, without limitation, vacancies, vacancy clusters, vacancy-impurity complexes, dislocations, grain boundaries, voids, and interfaces. Complete annihilation of a positron and an electron occurs when both particles collide and their combined mass is converted into energy in the form of two (and occasionally three) photons (e.g., gamma rays). If the positron and the electron are both at rest at the time of annihilation, the two gamma rays are emitted in exactly opposite directions (e.g., 180° apart) in order to satisfy the requirements of the conservation of momentum. Each annihilation gamma ray has an energy of about 511 keV, the rest energies of an electron and a positron. In positron annihilation techniques nearly all the positrons are at rest in the defect or lattice sites. However, the electrons are not. Therefore, the momentum of the electron tends to determine the momentum of the annihilating pairs and cause the direction of the gamma rays to deviate from 180°. In addition to the momentum constraints, the energies of the gamma rays resulting from the annihilation may deviate slightly from 511 keV, depending on the momentum of the electron. Accordingly, in non-destructive testing techniques utilizing positron annihilation, the detection of the energies and relative angles of the gamma rays produced by the annihilation event are used to derive certain information relating to defects and other characteristics of the material or object being tested. While positron annihilation techniques of the type described above have been successfully used in the laboratory to detect defects in specimen materials, the technique has not been successfully utilized in field settings. For one thing, the positrons from the external positron source barely penetrate the surface of the material being tested. Consequently, such external positron source techniques are limited to near surface measurements and generally must be conducted under controlled laboratory conditions. Partly in an effort to solve the depth limitations of the foregoing positron annihilation testing technique, another type of positron annihilation technique has been developed that replaces the external positron source with an external neutron source. Neutrons from the neutron source are directed toward the material being tested. Given sufficient energies, the neutrons will, in certain materials, result in the formation of isotopes that produce positrons. Such isotopes are commonly referred to as positron emitters. The positrons then migrate to lattice defect sites, ultimately annihilating with electrons to produce gamma rays. The resulting gamma rays are thereafter detected in the manner already described in order to derive information relating to the structure of the material being tested. The foregoing type of positron annihilation system is often referred to as a “neutron activated positron annihilation system” since it utilizes neutrons to trigger or induce the production of positrons. Since neutrons penetrate more deeply into the material being tested than do positrons alone (e.g., from an external positron source), such neutron activated positron annihilation systems are generally capable of detecting flaws deep within the material rather than merely on the surface. Unfortunately, however, only a relatively few elements, such as certain isotopes of copper, cobalt, and zinc, produce positrons in response to the neutron bombardment that are suitable for detecting flaws within the material. Consequently, neutron activated systems are limited to use with materials that contain such responsive elements. Non-destructive testing apparatus according to one embodiment of the invention comprises a photon source. The photon source produces photons having predetermined energies and directs the photons toward a specimen being tested. The photons from the photon source result in the creation of positrons within the specimen being tested. A detector positioned adjacent the specimen being tested detects gamma rays produced by annihilation of positrons with electrons which are indicative of a material characteristic of the specimen being tested. Also disclosed is a method that comprises the steps of: Providing a specimen having at least one positron emitter therein; determining a threshold energy for activating the positron emitter; and determining whether a half-life of the positron emitter is less than a selected half-life. If the half-life of the positron emitter is greater than or equal to the selected half-life, then activating the positron emitter by bombarding the specimen with photons having energies greater than the threshold energy and detecting gamma rays produced by annihilation of positrons in the specimen. If the half-life of the positron emitter is less then the selected half-life, then alternately activating the positron emitter by bombarding the specimen with photons having energies greater then the threshold energy and detecting gamma rays produced by positron annihilation within the specimen. Non-destructive testing apparatus 10 according to one embodiment of the present invention is illustrated in FIG. 1 and may comprise a photon source 12 and a detector 14. The photon source 12 produces photons (illustrated schematically by arrow 16) and directs the photons 16 toward a material or specimen 18 being tested. The photons 16 from the photon source 12 activate positron emitters (not shown) within the material or specimen 18, resulting in the creation of positrons (also not shown). Many of the positrons so formed ultimately annihilate with electrons (not shown) within the specimen 18, resulting in the formation of gamma rays (illustrated schematically by arrow 20). The gamma rays 20 resulting from the positron annihilations occurring within the specimen 18 are detected by the detector 14 which produces raw data 22 related to the detected gamma rays 20. A data collection and processing system 24 operatively associated with the detector 14 is responsive to the raw data 22 produced by the detector 14 and processes the raw data 22 to produce output data 26 that are indicative of at least one material characteristic of the specimen 18 being tested. Thereafter, the output data 26 may be presented in human-readable form an a suitable display system 28. As will be described in greater detail below, the method and apparatus of the present invention are suitable for use with materials or specimens 18 that will produce positrons in response to photon bombardment from the photon source 12. One way for producing positrons involves the decay of neutron-deficient isotopes. In the present invention, the photons 16 from the photon source 12 produce such neutron-deficient isotopes within the specimen 18 by removing or “knockingoff” neutrons from atoms within the specimen 18. The neutron-deficient isotopes (referred to herein in the alternative as “positron emitters”) then decay into nonneutron-deficient atoms by the emission of positrons and neutrinos. Consequently, the bombardment of a material or specimen 18 containing certain isotopes amenable to the loss of neutrons by such photon bombardment will result in the formation of positrons within the material or specimen 18. This process is referred to herein as “photoneutron activation” or, simply, “photon activation.” Any material containing isotopes susceptible to such photon activation is suitable for use with the present invention. A method 30 illustrated in FIG. 2 may be used to determine at least one material characteristic of the object or specimen 18. The first step 32 in the method 30 involves determining whether the material or specimen 18 to be analyzed includes one or more isotopes or “positron emitters” that are capable of photon activation. Stated another way, step 32 is used to identify those isotopes contained in the specimen 18 that will produce positrons in response to photon bombardment. A next step 34 in the method 30 determines the photon energy required to activate at least one of the isotopes or positron emitters identified in step 32. As will be described in greater detail below, the method and apparatus of the present invention allow a user to select for activation certain ones of the isotopes or positron emitters comprising the specimen 18. Accordingly, certain isotopes within the specimen 18 may be activated, while leaving other isotopes un-activated. The ability to selectively activate certain positron emitters will allow a user to determine several material characteristics of the specimen 18, including, for example, the amount or quantity of the selected positron emitters present in the specimen 18 as well as the locations of such positron emitters. Such information may be useful in ascertaining a wide range of material characteristics of the specimen 18, as will be described in greater detail below. Step 36 of the method 30 assesses the half-life of the photon activated isotope or positron emitter to be activated. If the half-life of the positron emitter is greater than a certain time (e.g., typically a few minutes or greater), then the method 30 utilizes a normal activation/analysis process 38 to test the specimen 18. Alternatively, if the half-life of the positron emitter is less than the certain time (e.g., typically on the order of tens of seconds or less), the specimen 18 is tested or analyzed in accordance with a rapid activation/analysis process 40. The normal activation/analysis process 38 is best seen in FIG. 3. A first step 42 in the normal activation/analysis process 38 involves activating the positron emitter or emitters (i.e., the isotope or isotopes identified in step 32). In one preferred embodiment, the positron emitter is activated by bombarding the specimen 18 with photons 16 from the photon source 12. It is generally preferred that the photons 16 from the photon source 12 have sufficient energies to activate the selected isotope or positron emitter. For example, and as will be described in greater detail below, photons having energies in the range of about 8 million electron volts (MeV) to about 22 MeV will activate most of the isotopes (i.e., positron emitters) likely to be found in many common materials. See, for example, Tables I and II. Alternatively, of course, photons having energies either above or below this range may be used, depending on the particular isotope and on the particular material characteristics to be detected. The photon-activated positron emitters result in the production of positrons within the specimen 18. Such positrons diffuse or migrate through the material comprising specimen 18 and tend to be attracted to voids or other lattice defects having favorable electronic potentials. Ultimately, a significant number of positrons will annihilate with electrons, resulting in the formation of gamma rays 20. Such gamma rays 20 are detected in step 44 by the detector 14, which produces raw data 22. The raw data 22 are then analyzed in step 46 to produce output data 26 that are indicative of at least one material characteristic of the specimen 18. The raw output data 26 may be displayed in suitable form on the display system 28. See FIG. 1. If the half life of the isotope or positron emitter to be activated is less than the certain time (e.g., typically on the order of tens of seconds or less), as determined in step 36 (FIG. 2), the method 30 executes the rapid activation/analysis process 40. With reference now to FIG. 4, the rapid activation/analysis process 40 involves alternate photon bombardment and subsequent gamma ray detection of the specimen 18. More specifically, the specimen 18 is first exposed to the photons 16 from the photon source 12 for a selected time (e.g., 10 minutes) at step 48. That is, the positron emitter or emitters are activated. Then, gamma rays 20 resulting from the annihilation of positrons with electrons are detected via detector 14 at step 50. If a sufficient number of gamma rays 20 have been detected, as determined in step 53, the method 30 proceeds to step 54 wherein the data are analyzed to produce output data 26 (FIG. 1) that are indicative of at least one material characteristic of the specimen 18. The output data 26 may be displayed in suitable form on the display system 28. Alternatively, if an adequate number of gamma rays 20 have not been detected, as determined in step 53, the method 30 returns to step 48 wherein the specimen 18 is again exposed to photons 16 from the photon source 12 for the selected time. That is, the positron emitters comprising the specimen 18 are re-activated. The activation and detection steps 48 and 50 are repeated until a sufficient number of gamma rays 20 have been detected. The alternate photon activation and detection steps 48 and 50, respectively, may be accomplished in a variety of ways. For example, in one preferred embodiment, the specimen 18 is alternately moved between an activation position 56 and a detection position 58. See FIG. 5. While in the activation position 56, the specimen 18 is positioned adjacent the photon source 12 so that the specimen receives photons 16 therefrom. Then, after having been exposed to the photons 16 for the selected time, the specimen 18 is moved to the detection position 58. While in the detection position 58, the detector 14 detects gamma rays 20 emitted from the specimen 18 as a result of positron/electron annihilations. However, other arrangements are possible for accomplishing the activation and detection steps 48 and 50. For example, in an alternative arrangement, the photon source 12 is alternately energized for the selected time period, then de-energized for a detection time period in which gamma rays 20 emitted from the specimen 18 are detected by the detector 14. Referring now to FIGS. 1 and 6, the data collection and processing system 24 may be provided with a data processing system 60 which processes the raw data 22 from the detector 14 in accordance with one or more algorithms in order to produce the output data 26 which are indicative of at least one material characteristic of the specimen 18. For example, in one preferred embodiment, the data processing system 60 may process the data 22 in accordance with a Doppler broadening algorithm 62, a positron lifetime algorithm 64, and a three-dimensional (3-D) imaging algorithm 66. The various algorithms (e.g., 62, 64, and 66) process the data 22 from the detector 14 in order to produce output data 26 which are indicative of at least one material characteristic of the specimen 18. For example, the Doppler broadening algorithm 62 is useful in assessing the characteristics of lattice defects contained in the specimen 18, such as, for example, damage resulting from mechanical and thermal fatigue, embrittlement, annealing, or manufacturing defects. The positron lifetime algorithm 64 is also useful in assessing the characteristics of lattice defects. In addition, information obtained from the mean lifetime of various defect components may be used to derive information relating to changing characteristics of the defects present in the specimen 18. The 3-D imaging algorithm 66 may be used to in conjunction with either the Doppler broadening algorithm 62 or the positron lifetime algorithm 64 to produce three-dimensional information regarding locations of the lattice defects contained within the specimen 18. Alternatively, the raw gamma ray data 22 from the detector 14 may be processed in accordance with other algorithms that are now known in the art or that may be developed in the future to derive other types of information, as would be obvious to persons having ordinary skill in the art after having become familiar with the teachings of the present invention. Consequently, the present invention should not be regarded as limited to the particular processing algorithms shown and described herein. Regardless of the particular algorithm (e.g., 62, 64, or 66) that is used to process the raw data 22, the resulting output data 26 may be presented in human-readable form on a suitable display system 28, such as a CRT or LCD display. Alternatively, other types of display systems may be used to present the output data 26 in useable form. For each algorithm, e.g., 62, 64, and 66, the data processing system 60 may utilize a selective activation algorithm 68 in which certain isotopes or positron emitters in the specimen 18 are selected to be activated. Stated simply, the selective activation algorithm 68 allows the data processing system 60 to set the energy level of the photons 16 produced by the photon source 12. See FIG. 1. As mentioned above, the selective activation algorithm 68 provides the option to allow the user to activate certain of the isotopes or positron emitters comprising the specimen 18. A significant advantage of the present invention is that since the positrons are produced within the material or specimen itself, rather than externally, the method and apparatus of the present invention may be used to determine material characteristics of the specimen 18 throughout the thickness (i.e., depth) of the specimen 18. Another significant advantage of the present invention is that it may be used in conjunction with a wide range of materials, including metals, polymers, and composite materials, that were not heretofore available for full depth study by positron annihilation methods. Still yet other advantages are associated with the ability to produce the positrons within the material specimen itself. For example, the invention realizes increased sensitivity over conventional positron annihilation methods utilizing external positron sources in that there is no extraneous background “noise” caused by annihilations external to the specimen being analyzed. The increased sensitivity also allows other types of detectors (e.g., CdZnTe) to be used. Moreover, the surface of the specimen need not be specially prepared as is typically required with systems involving external positron sources. The analysis techniques herein are also primarily dependent on and sensitive to the atomic characteristics of the specimen 18 and are not dependent on the physical geometry of the specimen. Another significant advantage of the present invention is that it may be made specific to particular isotopes within the specimen. That is, by adjusting the energies of the photons 16 from the photon source 12, the photons 16 may be used to selectively activate one or more positron emitters within the specimen 18 while leaving other positron emitters unactivated. Moreover, compared with conventional positron annihilation analysis devices, the present invention may be made quite small and portable, thereby allowing the present invention to be readily and easily utilized in field settings to analyze materials and specimens in-situ. The present invention may also be used to monitor materials during production and/or processing, thereby allowing for the early detection of non-compliant materials and for the possibility of adjusting production parameters and processes to minimize the creation of non-compliant materials. With the foregoing considerations in mind, non-destructive testing apparatus 10 according to one embodiment of the present invention is best seen in FIG. 1 and may comprise a photon source 12 and a detector 14. The photon source 12 produces photons 16 and directs the photons 16 toward the specimen 18 being tested. It is generally preferred, but not required, that the photon source 12 be capable of producing photons 16 having user adjustable (i.e. selectable) energies. The ability to adjust or select the energy of the photons 16 allows a user, in certain situations, to selectively activate only certain ones of positron emitters or isotopes (not shown) comprising specimen 18 while leaving certain other positron emitters un-activated. Alternatively, if such selective activation of the positron emitters is not required or desired in a particular application, the photon source 12 need not be provided with capability to adjust the photon energy. In one preferred embodiment having the ability to select the energies of the photons 16, the photon source 12 may comprise an electron accelerator 70 for producing a stream of accelerated electrons, shown schematically in FIG. 1 as broken line 72. In order to produce the photons 16 used to bombard the specimen 18, the accelerated electrons 72 are directed toward a target 74 which produces the photons 16 in response to bombardment by the accelerated electron stream 72. Photons generated in this manner are often referred to in the art as bremsstrahlung photons. There is a correlation between the energies of the electrons comprising the electron stream 72 and the photons produced by the target 74 in response to the electron bombardment. Consequently, photons 16 having specified energies can be produced by selecting or adjusting the energies of the electrons contained in the electron stream 72. In the embodiment shown and described herein, the photons 16 produced by the photon source 12 may be selected to have energies in the range of about 8 million electron volts (MeV) to about 22 MeV. Photons 16 having energies in this range are often referred to as gamma rays. In accordance with the foregoing considerations, then, the electron accelerator 70 may comprise a linear accelerator of the type that are now known in the art or that may be developed in the future that would be suitable for the production of electrons in any of a wide range of energies. By way of example, in one preferred embodiment, the electron accelerator 70 comprises a model 6000 linear accelerator available from Varian Corp. of Palo Alto, Calif. Alternatively, equivalent devices from the same or other manufacturers may also be used. The target 74 which produces the photons 16 may comprise tungsten, although other materials may also be used. Of course, the photon source 12 and/or the various components comprising the photon source 12 (e.g., the electron accelerator 70 and target 74) may be provided with suitable shielding materials (not shown), to prevent the unwanted escape of radiation from the photon source 12. In another embodiment of the invention, the photon source 12 may comprise a radioactive isotope (not shown) suitable for producing gamma radiation having sufficient energies to activate at least one positron emitter contained in the specimen 18 to be tested. While the use of such an isotopic gamma ray source has the advantage of dispensing with the need for an electron accelerator and target, most isotopic gamma ray sources do not readily lend themselves to producing gamma rays having energies that can be selected and varied by the user. However, the gamma rays produced by certain isotopic sources do have known and generally predictable energies, thus would be suitable for activating positron emitters having threshold (i.e., activation) energies generally at or below the energies of the gamma rays produced by the isotopic gamma ray source. The detector apparatus 14 may be positioned adjacent the photon source 12 and the specimen 18 so that the detector 14 receives gamma rays 20 resulting from positron/electron annihilation events occurring within the specimen 18. Depending on the geometry of the particular installation, a shield 76 may be positioned between the photon source 12 and the detector 14 to prevent gamma radiation from the photon source 12 from being detected by detector 14. The detector 14 may be provided with a collimator 78 to collimate the gamma rays 20. The detector 14 may comprise any of a wide range of gamma ray detectors that are now known in the art or that may be developed in the future that are or would be suitable for detecting gamma rays 20 produced by the annihilation of positrons and electrons within the specimen 18. Accordingly, the present invention should not be regarded as limited to any particular type of gamma ray detector. However, by way of example, in one preferred embodiment, the detector 14 may comprise germanium detector of the type that is well-known in the art and readily commercially available. Alternatively, the detector 14 could comprise a cadmium-zinc-tellurium (CdZnTe) detector of the type that is also well-known in the art and readily commercially available. The collimator 78 may comprise a variable slit type or other collimator. It should also be noted that the present invention is not to be regarded as limited to use with only a single detector. Indeed, many of the algorithms utilized by the present invention require, or at least prefer, the use of more than one detector. For example, the positron lifetime algorithm 64 will generally require the use of at least two detectors, one to detect the gamma rays 20 resulting from the annihilation events and one to detect “precursor” radiation associated with the production of the positrons themselves. Similarly, the 3-D imaging algorithm 66 will also generally utilize at least two, and preferably several, gamma ray detectors 14 in order to determine the position of the positron/electron annihilation event within the specimen 18. However, since the positron lifetime techniques and 3-D imaging techniques are well-known in the art, as are the requirements for the particular types and positions of detectors associated with such techniques, and since such multiple detectors could be easily provided by persons having ordinary skill in the art after having become familiar with the teachings of the present invention, the particular configurations of such multiple detector systems as they could be utilized in the present invention will not be described in further detail herein. The data acquisition and processing system 24 is operatively associated with the detector apparatus 14 and receives raw data 22 from the detector apparatus 14. In the embodiment shown and described herein, the data acquisition and processing system 24 may comprise a data acquisition system 80, as well as the data processing system 60. The data acquisition system acquires the raw data 22 from the detector and converts it into a form suitable for use by the data processing system 60. For example, in the case where the data processing system 60 comprises a digital computer system, the data acquisition system 80 may include an analog-to-digital (A/D) converter (not shown) suitable for converting the analog data 22 from the detector 14 into digital data suitable for use by the data processing system 60. Of course, other arrangements and configurations are possible, as would be obvious to persons having ordinary skill in the art after having become familiar with the teachings of the present invention. Consequently, the present invention should not be regarded as limited to any particular type of data acquisition system 80. However, by way of example, in one preferred embodiment, the data acquisition system 80 comprises a digital data acquisition system available from EG&G of Oak Ridge, Tenn. as model no. “DSPEC+”. Alternatively, similar systems from the same or other manufacturers may also be used. Initial analysis may be performed using the “Gamma Vision” software package commercially available from EG&G or the “Genie 2000” software package commercially available from Canberra of Meriden, Conn. As will be described in greater detail below, in-depth analysis is performed using algorithms to assess peak shape characteristics including shape comparisons, width ratios, and other shape characteristics. The data processing system 60 may comprise a general purpose programmable digital computer, such as the ubiquitous personal computer, configured to operate in the manner described herein. Alternatively, the data processing system 60 may comprise an application specific computer that is customized to operate in accordance with the teachings herein. Regardless of the particular type of system that is used, the data processing system 60 receives data from the data acquisition system 80 and processes it in order to produce output data 26. The output data 26 may be presented in human-readable form on any of a wide range of devices or systems, such as the display system 28, in order to indicate for the user at least one material characteristic of the specimen 18 being analyzed. By way of example, in one preferred embodiment, the display system 28 may comprise a color display system (such as a CRT or LCD display) that is operatively associated with the data processing system 60. Alternatively, other systems may be used, as would be obvious to persons having ordinary skill in the art. It is generally preferred, but not required, that the data processing system 60 also be operatively associated with the photon source 12. Such an arrangement allows the data processing system 60 to control the function and operation of the photon source 12, such as, for example, to select the desired photon energy, as well as to activate and deactivate the photon source 12, as may be required by the rapid activation/analysis process 40 (FIGS. 2 and 4) that may be utilized by the method 30 of the present invention. Alternatively, of course, such systems integration need not be provided. For example, the operation of the photon source 12 instead could be manually controlled by the user. Before proceeding with the description, it should be noted that the method and apparatus of the present invention may be used with materials or specimens 18 that will produce positrons in response to photon bombardment from the photons 16 produced by photon source 12. That is, the specimen 18 should comprise at least one positron emitter that, when “activated,” results in the production of positrons within the specimen 18. As mentioned above, one way for generating positrons is through the formation within the specimen 18 of neutron-deficient isotopes, i.e., positron emitters. Such neutron-deficient isotopes generally decay via the emission of positrons and neutrinos. A list of positron emitters, the threshold gamma ray energies required to form or “activate” the positron emitters, as well as their half-lives are presented herein as Tables I and II. Table I includes those isotopes having half-lives on the order of minutes or longer, whereas Table II includes short-lived isotopes having half-lives on the order of tens of seconds or less. It is generally preferred that such short-lived isotopes (i.e., the isotopes listed in Table II) be analyzed with the rapid activation/analysis process 40 shown and described herein. Tables I and II may be used to readily identify those isotopes that may be converted into positron emitters by photon bombardment as well as to estimate the photon energy required to form the positron emitters. TABLE IPositron EmittersThresholdElementReactionHalf-LifeUnitsEnergy MeVChromium50Cr → 49Cr42.3Minutes20.5Iron54Fe → 53Fe8.51Minutes14Nickel58Ni → 57Ni35.6Hours12Copper65Cu → 64Cu12.7Hours8Copper63Cu → 62Cu9.74Minutes11Zinc64Zn → 63Zn38.5Minutes20.45Zirconium90Zr → 89Zr4.18Minutes12.3Molybdenum92Mo → 91Mo1.08,Minutes12.515.5Tin112Sn → 111Sn35Minutes12.5Antimony121Sb → 120Sb15.9Minutes10Titanium46Ti → 45Ti3.1Hours13Carbon12C → 11C20.3Minutes19Nitrogen14N → 13N9.97Minutes10.5Oxygen15O → 14O122.2SecondsNDFluorine19F → 18F1.83Hours20Phosphorus31P → 30P2.5Minutes10.9Chlorine35Ci → 34Ci32.2MinutesNDPotassium39K → 38K7.6Minutes12.5Gallium69Ga → 68Ga1.13HoursNDSelenium74Se → 73Se40Minutes12Bromine79Br → 78Br6.45MinutesNDRuthenium96Ru → 95Ru1.64HoursNDPalladium102Pd → 101Pd8.4HoursNDSilver107Ag → 106Ag24Minutes9.0Cadmium106Cd → 105Cd55.5MinutesNDIndium113In → 112In14.4MinutesNDXenon124Xe → 123Xe2HoursNDCerium136Ce → 135Ce17.7MinutesNDPraseodymium141Pr → 140Pr40Minutes7Neodymium142Nd → 141Nd1.04Minutes9.5Samarium144Sm → 143Sm8.83Minutes12.5Europium151Eu → 150Eu12.8HoursNDErbium164Er → 163Er1.25HoursND TABLE IIShort Half-Life Positron EmittersThresholdElementReactionHalf-LifeUnitsEnergy MevNeon20Ne → 19Ne17.2SecondsNDMagnesium24Mg → 23Mg11.32Seconds16Aluminum27Al → 26Al6.3SecondsNDSilicon28Si → 27Si4.14SecondsNDSulfur32S → 31S2.56Seconds15Argon36Ar → 35Ar1.77SecondsND With reference now to FIG. 2, the method 30 of the present invention may be used to determine at least one material characteristic of the specimen 18. The first step 32 in the method 30 comprises determining whether the material or specimen 18 to be analyzed includes one or more isotopes or “positron emitters” that are capable of photon activation. That is, step 32 involves a determination of the positron emitter or emitters to be activated. Tables I and II may be used for this purpose. For example, if it is known that the specimen 18 contains 50Cr, photons 16 having sufficient energy may be used to produce or form 49Cr, a positron emitter. The next step 34 in the method 30 involves a determination of the photon energy required to activate at least one of the isotopes or positron emitters identified in step 32. For example, 50Cr has a threshold energy of 20.5 MeV. Therefore, photons 16 having energies greater than or equal to this value will interact with 50Cr to produce the positron emitter 49Cr. Of course, photons 16 having energies sufficient to activate chromium-50 will also activate other positron emitters contained in the specimen 18 having lower threshold energies. Step 36 of the method 30 assesses the half-life of the selected photon activated isotope(s) or positron emitter(s). In this regard it should be noted that if the half-life of the positron emitter is greater than a certain time (e.g., generally on the order of minutes or longer), then it will be advantageous to utilize the normal activation/analysis process 38 to test the specimen 18. Alternatively, if the half-life of the positron emitter is less than the certain time (e.g., on the order of tens of seconds or less), the specimen 18 may be tested or analyzed in accordance with the rapid activation/analysis process 40. In the example discussed herein involving chromium, Table I indicates that the half-life of the positron emitter 49Cr is about 42.3 minutes. Therefore, it will be preferable to utilize the normal activation/analysis process 38 for this positron emitter. The normal activation/analysis process 38 is best seen in FIG. 3. The first step 42 in the normal activation/analysis process 38 involves activating the positron emitter (i.e., the isotope or isotopes identified in step 32). In one preferred embodiment, the positron emitter is activated by bombarding the specimen 18 with photons 16 from the photon source 12 having energies sufficient to activate the selected positron emitter or emitters, as the case may be. As mentioned above, photons having energies in the range of about 8 MeV to about 22 MeV will activate most of the isotopes (i.e., positron emitters) likely to be found in many common materials. See, for example, Tables I and II. Alternatively, of course, photons having energies either above or below this range may be used, depending on the particular isotope and on the particular material characteristics to be detected. In the example involving chromium-49, the photons 16 produced by the photon source 12 should have energies of at least 20.5 MeV. The photon-activated positron emitters result in the production of positrons within the specimen 18. Such positrons diffuse or migrate through the material comprising specimen 18 and tend to be attracted to voids or other lattice defects having a favorable electronic potential. Ultimately, a significant number of the positrons produced by the positron emitter or emitters will annihilate with electrons, resulting in the formation of gamma rays 20. Such gamma rays 20 are detected in step 44 by the detector 14, which produces raw data 22. The raw data 22 are then analyzed in step 46 to produce output data 26 indicative of at least one material characteristic of the specimen 18. The output data 26 may be displayed in suitable form on the display system 28. See FIG. 1. If the half life of the isotope or positron emitter to be activated is less than a few tens of seconds, as determined in step 36, the method 30 executes the rapid activation/analysis process 40. With reference now to FIG. 4, the rapid activation/analysis process 40 involves alternate photon bombardment and subsequent gamma ray detection of the specimen 18. More specifically, the specimen 18 is first exposed to the photons 16 from the photon source 12 for a selected time at step 48. Then, gamma rays 20 resulting from the annihilation of positrons with electrons are detected via detector 14 at step 50. If a sufficient number of gamma rays 20 have been detected, as determined in step 53, the method 30 proceeds to step 54 wherein the data are analyzed to produce output data 26 (FIG. 1) that are indicative of at least one material characteristic of the specimen 18. The output data 26 may be displayed in suitable form on the display system 28. Alternatively, if an adequate number of gamma rays 20 have not been detected, the method 30 returns to step 48 wherein the specimen 18 is again exposed to photons 16 from the photon source 12 for a selected time. This rapid activation/analysis process 40 is repeated until a sufficient number of gamma rays 20 have been detected. The alternate photon activation and detection steps 48 and 50, respectively, may be accomplished in a variety of ways. For example, with reference now to FIG. 5, the specimen 18 could be alternately moved between an activation position 56 and a detection position 58. A suitable mechanical arrangement (not shown) may be provided to move the specimen 18 between the activation position 56 and the detection position 58. Alternatively, of course, the specimen 18 could remain stationary while the photon source 12 and detector 14 are moved. Again, a suitable arrangement for so moving the photon source 12 and detector 14 could be easily arrived at by persons having ordinary skill in the art after having become familiar with the teachings of the present invention. Regardless of the particular arrangement for moving the specimen 18 between the activation position 56 and the detection position 58 (or for moving the photon source 12 and detector 14), the specimen 18, while in the activation position 56, is positioned adjacent the photon source 12 so that the specimen 18 receives photons 16 therefrom. Then, after having been exposed to the photons 16 for the selected time, the specimen 18 is moved to the detection position 58. While in the detection position 58, the detector 14 detects gamma rays 20 emitted from the specimen 18 as a result of positron/electron annihilations. The times in which the specimen 18 is located in the activation position 56 and in the detection position 58 will vary depending on the particular positron emitter or emitters involved and on the particular material characteristics to be studied. However, the time during which the specimen 18 remains in the activation position 56 should be sufficient to activate a sufficient number of positron emitters so that the gamma rays 20 resulting from positron/electron annihilations will be detectable by the detector 14. Similarly, the specimen 18 should remain in the detection position 58 for a time sufficient to detect gamma rays 20 resulting from annihilation events. Generally speaking, the time that the specimen 18 should remain in the detection position 58 should be at least equal to one half-life of the activated positron emitter or emitters, although the time could be longer or shorter than the half-life. In consideration of these matters, then, the present invention should not be regarded as limited to any particular times for each position. As was briefly mentioned above, other arrangements are possible for alternately activating the positron emitters then detecting the gamma rays 20 resulting from annihilation events. For example, in another arrangement, the photon source 12 is alternately energized for the activation time period, then de-energized for a detection time period in which gamma rays 20 emitted from the specimen 18 are detected by the detector 14. Again, the activation time period should be set so as to activate a sufficient quantity of positron emitters, whereas the detection time period should encompass at least one half-life of the activated positron emitter or emitters. The data collection and processing system 24 may be provided with a data processing system 60 which may process the data 22 from the detector 14 in accordance with one or more algorithms in order to produce the output data 26 which are indicative of at least one material characteristic of the specimen 18. For example, with reference now to FIG. 6, in one preferred embodiment, the data processing system 60 may process the data 22 in accordance with a Doppler broadening algorithm 62, a positron lifetime algorithm 64, and a three-dimensional (3-D) imaging algorithm 66. The various algorithms (e.g., 62, 64, and 66) process the data 22 from the detector 14 in order to produce output data 26 which are indicative of at least one material characteristic of the specimen 18. The Doppler broadening algorithm 62 is useful in assessing the characteristics of lattice defects contained in the specimen 18. Such lattice defects may include, without limitation, damage resulting from mechanical and thermal fatigue, embrittlement, annealing, and manufacturing defects. Doppler broadening techniques involve an assessment of the degree of broadening of the 511 keV peak associated with the gamma rays 20 produced by the positron/electron annihilation event. Basically, a broadening of the peak is indicative of the presence of one or more lattice defects. Several different types of Doppler broadening techniques have been developed and are being used in the positron annihilation art and could be easily implemented in the present invention by persons having ordinary skill in the art after having become familiar with the teachings of the present invention. Accordingly, the present invention should not be regarded as limited to any particular type of Doppler broadening technique. However, by way of example, in one preferred embodiment of the invention, the Doppler broadening algorithm 62 may comprise the Doppler broadening algorithm described in U.S. Pat. No. 6,178,218 B1, which is specifically incorporated herein by reference for all that it discloses. The positron lifetime algorithm 64 is also useful in assessing the characteristics of lattice defects. For example, the positron lifetime algorithm 64 may be used to obtain information as to whether the lattice defects comprise monovacancies, dislocations, slip zones, or particulate inclusions. In addition, information obtained from the mean lifetime of various defect components may be used to derive information relating to changing characteristics of the defects present in the specimen. The positron lifetime algorithm 64 basically involves a determination of the time between positron formation and positron annihilation. In order to do so, the positron lifetime algorithm detects some precursor event associated with the formation of the positron, as well as the gamma rays 20 produced by the positron annihilation event. The time between these two events is the positron lifetime. In accordance with the foregoing process, systems utilizing positron lifetime analysis techniques usually utilize two separate detectors, one for detecting the precursor event and the other for detecting the annihilation event. The system will also usually include constant fraction discriminators, a time amplitude converter, as well as, a multi-channel analyzer system. However, since systems for detecting positron lifetimes, as well as the algorithms utilized thereby, are well-known in the art and could be easily provided by persons having ordinary skill in the art after having become familiar with the details of the present invention, the positron lifetime algorithm 64, as well as the other systems and detectors that may be required or desired, will not be described in further detail herein. The 3-D imaging algorithm 66 may be used to in conjunction with either the Doppler broadening algorithm 62 or the positron lifetime algorithm 64 to produce three-dimensional information regarding locations of the lattice defects contained within the specimen 18. That is, in addition to determining the presence and characteristics of lattice defects (e.g., which may be accomplished by either the Doppler broadening algorithm 62 or the positron lifetime algorithm 64), the 3-D imaging algorithm 66 is also able to determine the position within the specimen 18 of the lattice defects. Consequently, the 3-D imaging algorithm 66 is capable of providing a wealth of information regarding the internal structure of the specimen 18. As mentioned above, the 3-D imaging algorithm 66 will benefit from the use of two or more separate detectors (e.g., detectors 14) in order to accurately define the locations of the positron annihilation events. However, three dimensional imaging techniques of the type that may be utilized in the present invention, as well as multiple detector arrangements for the use of the same, are also well-known in the art and could be readily provided by persons having ordinary skill in the art after having become familiar with the teachings of the present invention. For example, any of the imaging techniques and detector arrangements that are currently utilized in positron emission tomography (PET) may be readily adapted for use with the present invention. Therefore, the particular 3-D imaging algorithm 66 (and detector arrangements) that may be utilized in one embodiment of the present invention will not be described in further detail herein. For each analysis algorithm, e.g., 62, 64, and 66, described above the data processing system 60 may utilize a selective activation algorithm 68. The selective activation algorithm 68 allows certain isotopes or positron emitters in the specimen 18 to be activated. The selective activation algorithm 68 is responsive to input from the user regarding either the particular positron emitter or emitters to be activated or the desired photon energy. The selective activation algorithm 68 then controls or operates the photon source 12 as necessary to produce photons 16 having energy levels suitable for activating the selected positron emitter or emitters. The selective activation algorithm 68 allows the user to activate certain of the isotopes or positron emitters comprising the specimen 18. It is contemplated that the inventive concepts herein described may be variously otherwise embodied and it is intended that the appended claims be construed to include alternative embodiments of the invention except insofar as limited by the prior art. |
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claims | 1. A device configured to control an intensity of a transmission of radiation from a source at a first location on a first side of the device to a second location on a second side of the device, the device comprising:a fluid reservoir;a Magneto-Rheological fluid comprising magnetically conductive particles that reflect neutrons, the Magneto-Rheological fluid occupying at least a portion of an interior of the fluid reservoir;a magnetic field generator for establishing a magnetic field across the Magneto-Rheological fluid; anda control system for controlling the magnetic field generator to vary the magnetic field across the Magneto-Rheological fluid to change an albedo of the device in accordance with a demand signal. 2. The device of claim 1 wherein the device is a nuclear reflector supported around at least a portion of a circumference of a core of a nuclear reactor, the nuclear reflector having at least a first state and a second state wherein a strength of the magnetic field is varied by the control system to change the nuclear reflector between the first and second states to change the albedo of the device. 3. The device of claim 2 wherein the magnetic field is established substantially parallel to an axial direction of the core of the nuclear reactor. 4. The device of claim 2 wherein the magnetic field is a plurality of discrete magnetic fields spaced around a circumference of the core of the nuclear reactor. 5. The device of claim 4 wherein the plurality of discrete magnetic fields are respectively established by a plurality of independently operated electromagnets. 6. The device of claim 1 including a monitoring system connected to the control system, configured to monitor the radiation on the first side of the device and generate and communicate the demand signal to the control system to adjust the magnetic field to control a level of the radiation on the first side of the device to a preselected level. 7. The device of claim 2 wherein the demand signal is a power demand signal of the nuclear reactor. 8. The device of claim 2 wherein the demand signal controls a level of the nuclear reactions within the core of the nuclear reactor by changing the magnetic field. 9. The device of claim 8 wherein the change in the magnetic field is the primary mechanism for controlling the nuclear reactions within the core during normal operation of the nuclear reactor. 10. The device of claim 2 wherein if power is lost to the magnetic field generator the Magneto-Rheological fluid transitions to a state that shuts down the reactor. 11. The device of claim 2 wherein the device alters the radial power distribution of the core. 12. The device of claim 2 wherein the device alters the axial power distribution of the core. |
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abstract | The present invention provides passive safety equipment, comprising: a cooling part formed to cool a first fluid, which is emitted from a reactor coolant system or a steam generator, and a second fluid in a housing; and a circulation induction sprayer which is formed to spray the first fluid emitted from the reactor coolant system or the steam generator into the cooling part, has at least part thereof open to the inside of the housing such that the second fluid flows thereinto according to a drop in pressure caused by the spraying of the first fluid, and sprays the second fluid with the inflown first fluid. |
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048030406 | description | DETAILED DESCRIPTION OF THE INVENTION Reference will now be made to the present preferred embodiments of the invention, an example of which is illustrated in FIGS. 1 and 2. FIG. 1 illustrates the architecture of the failed fuel surveillance and diagnosis apparatus of the present invention. An artificial intelligence based "inference engine" 10 along with a factual knowledge base 12 comprise a knowledge system which is interfaced to the reactor's data acquisition system and control room instrumentation. The artificial intelligence based inference engine 10 will hereinafter be referred to as a judgmental knowledge base. The judgmental knowledge base, along with the factual knowledge base, comprise a knowledge system which may be used to emulate a reasoning task to interpert encoded knowledge of human experts stored therein. The reactor parameters which are used as input into the judgmental knowledge base 10 include readings from the primary coolant flow rate with a flow meter 14 and readings of the reactor power with a reactor power level detector 16. Preferably, these readings include two independent readings of the primary coolant flow rate through a first flow channel 17 and a second flow channel 18 and two independent readings of the reactor power through a first power channel 19 and a second power channel 20. The remaining input parameters originate within the DN monitoring station (DNMS). The DNMS 22 generates output signal 24, indicating the age of the delayed neutrons, and signal 26, indicating the ERA of a breached fuel element, to judgmental knowledge base 10. Preferably, the DNMS 22 is a multiple detector DN monitoring system, as the one disclosed in U.S. Pat. No. 4,415,524 issued to Kenny C. Gross et al. This system includes at least three DN activity detectors 28 and an ERA meter 30. In the prefered embodiment, which utilizes the ERA meter disclosed by the above-referenced patent, the DN age signal 24 and ERA signal 26 are generated by means of the three DN detectors in combination with a loop flow circuit 32 and pump 34. The pump 32 conveys coolant from reactor core through the loop 32 and back into the core. The three DN activity detectors 28 are placed proximate to the loop flow circuit 32. The DN activity detectors generate output signals to computer system 36, which communicates with factual knowledge base 12. The computer system 36 generates the DN age signal 24 and the ERA signal 26 by means of the equations disclosed in the above-reference patent. Factual knowledge base 12 contains factual data, which is available to the judgmental knowledge base 10 in the decision making process. Factual knowledge base 12 contains data relating to: the radioactive decay constants for the DN emitting fission products; isotopic fission yields; recoil correction factors; probabilities of neutron emmission; known flow delay times between successive DN detectors in the DNMS 22; equations for variations in transit times as a function of coolant flow rate; DN detector efficiencies and calibration factors; and any other relevant nuclear and system data. An operability validaton system 38 detects any malfunction in the components of DNMS system 22 and interfaces signals indicating a malfunction to the judgmental knowledge base 10. In the exemplary embodiment wherein the multiple detector DN monitoring system is utilized, detecting system 38 is comprised of a system which detects the flow in flow circuit 32, the status of the pump 34 and the temperature of the flow through the loop 32. The system includes a flow metering device 44 which measures the flow rate through loop 32 and thermocouples 46 which measure the temperature of the flow in the circuit. The system also includes a voltage meter 40 and a current meter 42 which measure the voltage and current of pump 34 respectively. Although the apparatus of the present invention has been described with reference to a multiple detector DN monitoring system, it will b readily apparent to those skilled in the art that other systems which measure the ERA may also be used along with the appropriate operatibilty validation system. The judgmental knowledge base 10 receives, as inputs, signals indicating the primary flow rate, the reactor power level, the delayed neutron age, the ERA, and the signals indicating the operability of the components in the DNMS from operability validation detector 38. Judgmental knowledge base 10 then implements an operability logic algorithm, which is illustrated in FIG. 2. Output from the judgmental knowledge base 10 is integrated with a display monitor 48 in control room 50 and then multiplexed back to data acquisition system 100 for archive backup storage. During operation with a breached element that gives a DN signal, the total age (i.e. sum of T.sub.tr and T.sub.h) that is output from the ERA meter 30 will be continuously monitored. If the age is increasing, a check will first be made to determine if T.sub.h is increasing. If so, the ERA value will be compared against a predetermined shutdown limit. That limit will replace the current administrative limit on DN-signal magnitude, and is expected to be far more conservative in limiting events that might challenge safety or radiological performance guidelines, while minimizing the possibilities of unnecessary reactor trips caused by events having no safety significance. If the computed ERA value exceeds the limit, an audible alarm 54 will be sounded and the operator will initiate a manual shutdown of the reactor. If T.sub.h is not increasing, then the sodium transport time is increasing. In this event, a check is first made with the two independent primary flow sigaals from flow channels 17 and 18. If it is determined that the flow through the core is not being changed, an alarm status is set and the check is made for a malfunction affecting flow control within the DNMS loop 22 itself. Signals employed for this comparison are the DNMS electromagnetic pump voltage and current from voltage meter 40 and current meter 42 respectively, the DNMS flow from flow meter 44 and the loop flow temperature from thermocouples 46. If it is determined that the indicated change in the T.sub.tr is attributed to a malfuction in the DNMS loop 22, then an attempt can be made to correct the problem during the time period provided for in the technical specifications, or the operator can initiate a manual shutdown. Finally, in the unlikely event that Ttr would be increasing while all primary and DNMS-loop signals indicate nominal readings, then this would be an indication of a possible formation of an assembly flow blockage. The reactor would then be scrammed. In the preferred embodiment of the present invention, an interactive terminal 62 is interfaced with display monitor 48 and judgmental knowledge base 10. The reactor operator is then provided with an interactive capability to manually query the status of any component of the system for operability validation. Thus, in this embodiment of the present invention, the system may be operated in a passive surveillance mode. Use of the present invention will reduce complexity and mitigate confusion in the reactor control room 50. It will minimize the possibility of human error or oversight, by providing automatic annunciation of discrepant signals or the incipiences of initiating faults. It also provides the reactor operator with a passive surveillance mode. This combination of automatic and manual systems reduces challenges to plant availability while allowing incorporation of the role of the operator in a manner of which most effectively augments the achievement of overall plant operability goals. In summary, diagnostic information made available from the present invention will be processed, compared against derived information from independent physical sensors, and presented to the reactor operator with the aid of the artificial intelligence-based surveillance and diagnosis system of the present invention. This apparatus, will be multiplexed to output devices in the reactor control room 50, will provide the operator with rapid identification (as much as ten minutes in advance of signals from the cover-gas monitoring system) of conditions that could lead to plant operational degradation, enabling him or her to terminate or avoid events which might challenge safety or radiological performance guidelines. The foregoing description of the preferred embodiments of the invention have been presented for purposes of illustration and description. They are not intended to be exhaustive or to limit the invention to the precise forms disclosed, and obviously many modifications and variations are possible in light of the above teachings. The embodiments were choosen and described in order to best explain the principles of the invention and its practical applications to thereby enable others skilled in the art to best utilize the invention in various embodiments and with various modifications as are suited to the particular use contemplated. It is intended that the scope of the invention be defined by the claims appended hereto. |
description | This application claims the conventional priority based on Japanese patent application serial No. 2006-001301, filed on Jan. 6, 2006, the disclosures of which are incorporated herein by reference. 1. Field of the Invention The present invention generally relates to an apparatus and method for evaluating equipment operability for evaluating the operability of equipment, and more specifically to an apparatus and method for evaluating equipment operability for evaluating the operability of equipment by determining the visibility and the readability of the target operation area of an equipment model arranged in a three-dimensional virtual space. 2. Description of the Related Art Conventionally, when an equipment layout is verified, for example, a template is arranged in the target operation area of the equipment, and it is determined whether or not the hand of an operator of the equipment can reach the target operation area, or the target operation area can be within the range of the view of the operator of the equipment. The target operation area of equipment refers to the area on the equipment where a part of the equipment to be operated and a part (for example, a flicker (blinking unit), a label and so on) to be visually identified by the operator during the operation of the equipment are set. However, in verifying the layout of the equipment by determining whether or not the hand of the operator of the equipment reaches the target operation area, or the target operation area can be within the range of the view of the operator of the equipment, the target operation area can be out of the range of the view because there are practically various physiques. Furthermore, when an operation is actually performed, the hand of an operator can hide the displayed target operation area. Although the view is not interfered with, the hand of the operator can hide the target operation area and the operator cannot confirm the target operation area. Additionally, the visibility and the readability of the target operation area can be varied by the size and the brightness of the displayed target operation area. Under the circumstances, the layout of equipment is verified using a human body model arranged in a three-dimensional virtual space. That is, the layout of the equipment model displayed on the view image is verified by arranging the human body model of a physique and the equipment model in a three-dimensional virtual space, and displaying a view image as an image viewed in the three-dimensional virtual space from the point of view of the human body model. A human body model is a copy of the shape of a human body. An equipment model is a copy of the shape of equipment. Furthermore, Japanese Patent Application Laid-open No. 10-240791 describes an equipment operability evaluation support apparatus for supporting the evaluation of the operability of equipment by arranging an equipment model taking the shape of equipment to be designed or equipment to be laid out in a simulation space, arranging a human body model with an operation of the equipment anticipated, and operating the human body model in the simulation space. Japanese Patent Application Laid-open No. 11-167451 describes a measurement system capable of measuring a difference, a position and so on even under a special condition that can interfere with the recognition of an image by preparing a human body model corresponding to the body type data of a person and data of a posture or operation, and regenerating the posture or operation of the person from the data in a virtual space. To confirm an object that interferes with the operation of equipment or the view of an operator using a human body model arranged in a three-dimensional virtual space, it is necessary to confirm the object by allowing the human body model to take an operating posture. Conventionally, to allow the human body model in the three-dimensional virtual space to take an operating posture, for example, the calculation of a posture is performed based on the forward kinematics and a predetermined joint angle, and a view image as an image viewed in the three-dimensional virtual space from the point of view of the human body model taking the calculated posture is displayed. However, in this case, each joint angle is set independently for each physique of a human body model, and the operating posture is calculated. That is, when the physique of a human body model is changed, it is necessary to reset each joint angle of a human body model to regenerate the operating posture. That is because the length of each part of each human body model is different for each physique and the joint angle of the human body model taking an operating posture is changed when the physique is changed. Accordingly, if the posture of the human body model is calculated using the data of the joint angle before changing the physique when the physique of a human body model taking an operating posture is conventionally changed, the range of the view of the human body model is shifted, and it is difficult to verify the layout of the equipment. For example, assume that a human body model 100 and an equipment model 101 are arranged in a three-dimensional virtual space as shown in FIG. 19A. FIG. 19A shows a card slot 102, a range 200 of the view of the human body model 100, and a point P indicating the target position of the right hand of the human body model 100 to be moved. The card slot 102 is a target operation part of the equipment model 101. The equipment model 101 is, for example, a model of an ATM. The human body model 100 takes a posture for performing an operation of, for example, inserting a card, and the right hand of the human body model 100 is in the position of the point P. The card slot 102 is in the range 200 of the view of the human body model 100. In this state, when the physique of the human body model 100 is changed, the right of the human body model 100 is displaced from the point P as shown in FIG. 19B. Additionally, the card slot 102 is displaced from the range 200 of the view of the human body model. Furthermore, conventionally there has been the problem that the optimum posture of the human body model for an operation cannot be maintained when the layout of equipment is changed. Additionally, there has conventionally been a well-known technology of calculating a posture corresponding to an operation by associating the posture of a human body model arranged in a three-dimensional virtual space with the target operation part of target equipment. However, this method can only be used for the target equipment. Therefore, it is difficult to verify a layout when another piece of equipment is used in this method. Furthermore, there has been a well-known function of displaying a view image of a human body model by allowing the human body model to take an operating posture in a three-dimensional virtual space. However, there is the problem with this function that it cannot be determined on the visibility and the readability of the target operation area of an equipment model displayed on the view image. Conventionally, there also has been the problem that the visibility or the readability for the target operation area of an equipment model in a series of operations required to attain an operation purpose of equipment cannot be determined. For example, as shown in FIG. 20A, the human body model 100 is allowed to take an operating posture of inserting a card into the card slot 102 of the equipment model 101, and then to automatically take a posture of an operating transaction selection part 105 of the equipment model 101 as shown in FIG. 20B. It is hard to determine the visibility and the readability on the target operation area of the equipment model in these series of operations performed by the human body model 100 based on the above-mentioned process. It is an object of the present invention to provide an apparatus for evaluating equipment operability which evaluates the operability of equipment from the point of view of the human body model taking a posture adapted to the target position even when the physique of a human body model is changed. It is another object of the present invention to provide a method for evaluating equipment operability for evaluating the operability of equipment from the point of view of the human body model taking a posture adapted to the target position even when the physique of a human body model is changed. The apparatus for evaluating equipment operability of the present invention evaluates operability of equipment. The apparatus comprises a physique management unit managing physique information about a human body model as a copy of a shape of a human body, a target management unit managing a human body part of the human body model and a target position to which the human body part is to be moved as target information, a posture calculation unit calculating an operating posture taken when the human body model is operating an equipment model as a copy of a shape of the equipment using the target position as a watch point of the human body model based on the physique information about the human body model managed by the physique management unit and the target information managed by the target management unit, a simulation unit arranging the equipment model and the human body model taking the operating posture calculated by the posture calculation unit in a three-dimensional virtual space, and displaying a view image as an image of the three-dimensional virtual space viewed from the point of view of the human body model taking the operating posture, and a view determination unit determining whether or not a visual identification target part is in the range of the view of the human body model using the information about the visual identification target part as the part of the equipment model to be visually confirmed by the human body model. Preferably, the apparatus of the present invention further comprises a first determination unit determining visibility of a visual identification target part of the equipment model displayed on the view image and readability of a character string associated with the visual identification target part based on the information about the visual identification target part, and a second determination unit determining the operability of the equipment based on a determination result by the first determination unit. Preferably, the apparatus of the present invention further comprises an operating posture management unit managing an order in which the human body model takes the operating posture and the target information for use in calculating the operating posture for each operating posture taken by the human body model. The posture calculation unit calculates the operating posture of the human body model in the order in which the human body model takes the operating posture managed by the operating posture management unit. The method for evaluating equipment operability of the present invention evaluates operability of equipment using an equipment operability evaluation apparatus. The method comprises managing physique information about a human body model as a copy of a shape of a human body, managing a human body part of the human body model and a target position to which the human body part is to be moved as target information, calculating an operating posture taken when the human body model is operating an equipment model as a copy of the shape of the equipment using the target position as a watch point of the human body model based on the physique information about the human body model and the target information, arranging the equipment model and the human body model taking the calculated operating posture in a three-dimensional virtual space, and displaying a view image as an image of the three-dimensional virtual space viewed from a point of view of the human body model taking the operating posture, and determining whether or not a visual identification target part is in a range of view of the human body model using information about the visual identification target part as a part of the equipment model to be visually confirmed by the human body model. The apparatus and method for evaluating equipment operability evaluation apparatus according to the present invention calculate the operating posture as a posture taken when a human body model operates an equipment model as a copy of the shape of equipment using a target position as a watch point of the human body model based on the physique information about the human body model and the target position to which a human body part is to be moved, then arrange the equipment model and the human body model taking the calculated operating posture in the three-dimensional virtual space, display a view image as an image of the three-dimensional virtual space viewed from the point of view of the human body model, and determine whether or not the visual identification target part as a part of the equipment model to be visually confirmed by the human body model are in the range of the view of the human body model. Thus, when the physique of the human body model is changed, the human body model is correspondingly and automatically allowed to take the operating posture adapted to the target position, and it can be determined whether or not the visual identification target part is in the range of the view of the human body model. Thus, the visibility and the readability of the target operation area of the equipment model displayed on the view image can be determined. Additionally, the operating posture can be calculated without setting each joint angle independently for each physique of the human body model. Therefore, the operating posture can be changed without displacing the range of the view of the human body model when the physique of the human body model taking a certain operating posture is changed, and the layout of equipment can be verified. Furthermore, the visibility and the readability on the target operation area of the equipment model in a series of operations of the equipment can be determined. That is, by allowing the human body model to take a certain operating posture, automatically allowing it to take the next operating posture, and based on the processes, the visibility and the readability of the target operation area of the equipment model in a series of operations can be determined. As described above, according to the present invention, an object interfering with the operation of equipment and the view can be confirmed by using a human body model arranged in the three-dimensional virtual space and allowing the human body model to take an operating posture. Therefore, according to the present invention, although the physique of a human body model is changed, the human body model can be automatically allowed to take an operating posture adapted to a target position, and the visibility and the readability on the visual identification target part on the equipment model viewed from the point of view of the human body model can be determined. Furthermore, when the layout of equipment is changed, the optimum posture of a human body model for an operation can be maintained, and the layout can be verified on various types of equipment. The apparatus for evaluating equipment operability according to the present invention can also be configured to manage the information about the target part to be operated by a human body model as associated with the target position, keep the relative position to the target position associated with the target part constant, and calculate the operating posture as a posture taken when the human body model operates the equipment model using the target position as a watch point of the human body model based on the physique information about the human body model, the target information, and the target part information. According to the present invention having the above-mentioned configuration, when the position of the target part is moved, the target position is changed such that the relative position to the target part can be kept constant. Therefore, although the layout of equipment is changed and the position of the target part has been moved, the human body model can be allowed to automatically take an operating posture adapted to the change of the layout of the equipment. Additionally, the apparatus for evaluating equipment operability according to the present invention manages the order in which the human body model takes an operating posture and the target information used in calculating an operating posture for each operating posture taken by the human body model, and calculates the operating posture of the human body model in the managed order in which the human body model takes an operating posture. Therefore, according to the present invention, the visibility and the readability on the visual identification target part of an equipment model in a series of operations required to attain the operation purpose of equipment can be determined. FIG. 1 shows an example of the structure of the apparatus for evaluating equipment operability according to the present invention. An equipment operability evaluation apparatus 1 shown in FIG. 1 is a processing device for evaluating the operability of equipment by determining the visibility and the readability of a target operation area of an equipment model in the view of a human body model arranged in a three-dimensional virtual space. The equipment operability evaluation apparatus 1 includes an input unit 11, a managing/setting unit 2, an operation unit 3, an evaluation unit 4, and an output unit 22. The input unit 11 inputs instruction information for the equipment operability evaluation apparatus 1 according to the specification input of the user of the equipment operability evaluation apparatus 1. The input unit 11 inputs as the instruction information, for example, the necessary information for the evaluation of the operability of equipment managed by the managing/setting unit 2. The managing/setting unit 2 sets and manages the necessary information for the evaluation of the operability of equipment. The managing/setting unit 2 includes a physique management unit 12, a target management unit 13, a target part management unit 14, a visual identification management unit 15, a character string setting unit 16, and an operation management unit 17. The operation unit 3 calculates the operating posture of a human body model, arranges an equipment model and the human body model having the calculated operating posture in a three-dimensional virtual space, and displays a view image as an image of the three-dimensional virtual space viewed from the point of view of the human body model. The operation unit 3 includes a posture calculation unit 18 described later, and a simulation unit 19. The evaluation unit 4 evaluates the operability of equipment. The evaluation unit 4 includes a view determination unit 20 and an operation determination unit 21 described later. An output unit 22 outputs an evaluation result of the operability of equipment. The output unit 22 outputs, for example, the recorded contents in an operation determination table 30 as an evaluation result of the operability of equipment. The physique management unit 12 includes a physique management table 23, and manages the physique information about the human body model arranged in the three-dimensional virtual space. The managed physique information is stored in the physique management table 23. The target management unit 13 includes a target management table 24, and manages target information. The managed target information is stored in the target management table 24. The target management unit 13 manages as target information, for example, the control point of the human body model in the three-dimensional virtual space and a target position to which the control point is to be moved. The control point corresponds to, for example, a human body part such as a right hand, a left hand and so on. The target position is an operation position on an equipment model. The target part management unit 14 includes a target part management table 25, and manages target part information. The managed target part information is stored in the target part management table 25. The target part management unit 14 manages as target part information, for example, a target part and a relative position of the target part to the target position. The target part is a part to be operated in the parts of equipment. Therefore, a target part is normally arranged close to a target position. A target part can be, for example, a card slot and so on. The visual identification management unit 15 includes a visual identification management table 26, and manages visual identification information. Visual identification information is information about a part to be visually confirmed by a human body model among the parts of an equipment model. The managed visual identification information is stored in the visual identification management table 26. The visual identification management unit 15 manages as the visual identification information, for example, the information about a visual identification target part to be visually confirmed by a human body model among the parts of an equipment model. That is, the visual identification management unit 15 is a visual identification management unit managing as visual identification information the information about a visual identification target part of an equipment model to be visually confirmed by a human body model. A visual identification target part can be, for example, a flicker, a label and so on. The character string setting unit 16 sets (stores) the attribute information about a character string associated with each visual identification target part in a character string setting table 27. A character string associated with a visual identification target part corresponds to, for example, a character string displayed on or applied to a visual identification target part, a character string displayed on or applied to an area around the visual identification target part and so on. The attribute information about the set character string can be, for example, the position, size, color, and character width and so on of a character string displayed on a visual identification target part. The operation management unit 17 includes an operation management table 28, and manages operation information. The managed operation information is stored in the operation management table 28. The operation information is information about a series of operations required to attain the purpose of an operation. The operation information includes, for example, an order of operating postures in which a human body model takes each operating posture and target information. That is, the operation management unit 17 is an operating posture management unit managing as operation information an order of operating postures in which a human body model takes each operating posture and target part information for use in calculation of operating posture for each operating posture taken by the human body model. The operation management unit 17 can also be configured to manage as operation information an order of operating postures in which a human body model takes each operating posture and target part information for each operating posture taken by the human body model. The operation management unit 17 can further be configured to manage as operation information an order of operating postures in which a human body model takes each operating posture, target information or target part information, and visual identification information for each operating posture taken by the human body model. In this case, the visual identification information corresponding to the operating posture can be the information about a visual identification target part to be watched by a human body model or a visual identification target part to be visually confirmed by the human body model that is used when the posture calculation unit 18 described later calculates a posture of the human body model. The visual identification information corresponding to an operating posture can be the information about a visual identification target part to be processed in the view determining process, the visual identification determining process, and the character recognizing process performed by the view determination unit 20. The posture calculation unit 18 calculates a posture to be taken by a human body model in a three-dimensional virtual space. For example, the posture calculation unit 18 calculates the posture of the human body model for allowing the control point of the human body model having the physique corresponding to the physique information to move to a target position and for allowing the human body model to watch the target position based on the physique information of the human body model in the physique management table 23, the operation information in the operation management table 28, and the target information in the target management table 24. A target to be watched by a human body model can be a visual identification target part. For example, the posture calculation unit 18 can also calculate the posture of the human body model for allowing the control point of the human body model having the physique corresponding to the physique information to move to a target position and for allowing the human body model to watch the target position and visually confirm a visual identification target part based on the physique information of the human body model in the physique management table 23, the operation information in the operation management table 28, the target information in the target management table 24, and the visual identification information in the visual identification management table 26. Furthermore, for example, the posture calculation unit 18 can also calculate the posture of the human body model for allowing the control point of the human body model having the physique corresponding to the physique information to move to a target position at a predetermined relative position from a target part and for allowing the human body model to watch the target position based on the physique information of the human body model in the physique management table 23, the operation information in the operation management table 28, the target part information in the target part management table 25, and the target information in the target management table 24. The target to be watched by a human body model can be a visual identification target part. For example, the posture calculation unit 18 can also calculate the posture of the human body model for allowing the control point of the human body model having the physique corresponding to the physique information to move to a target position at a predetermined relative position from a target part and for allowing the human body model to watch the target position and visually confirm a visual identification target part based on the physique information of the human body model in the physique management table 23, the operation information in the operation management table 28, the target information in the target management table 24, and the visual identification information in the visual identification management table 26. For example, when the position of a target part to be operated is changed, the posture calculation unit 18 changes the target position used in calculating a posture such that the relative position to the target part can be in a predetermined range, and calculates the posture for allowing the control point of a human body model to move to the changed target position. As the posture calculation unit 18 calculates the posture of the human body model based on the target part information, the optimum posture of a human body model can be maintained for an operation although the layout of equipment has been changed. The posture calculation unit 18 outputs, for example, the angles of the joints at the neck, the shoulder, the elbow, and the wrist, and the position of the eyes of a human body model as posture information. Practically, the posture calculation unit 18 performs the calculation of the posture in the process using the well-known inverse kinematics (IK). The inverse kinematics generally refers to the function of setting a hierarchical structure of a parent, a child, a grandchild and so on for an object such that upper child and parent objects are moved when the grandchild object are firstly moved. The process used by the posture calculation unit 18 in calculating the posture according to the present invention is not limited to the inverse kinematics. For example, the posture calculation unit 18 can calculate the posture using the process of the well-known dynamics and so on. The simulation unit 19 arranges the human body model having the operation posture calculated by the posture calculation unit 18 together with the equipment model in the three-dimensional virtual space. The simulation unit 19 displays a view image as an image of the three-dimensional space viewed from the point of view of the human body model as described later by referring to FIG. 15. For example, the simulation unit 19 generates an image of an equipment model when a predetermined light source emits light in a well-known process such as, for example, a radio city method and so on to the equipment model arranged in the three-dimensional virtual space as described later. The simulation unit 19 includes a human body model storage unit 31 and an equipment model storage unit 32. The human body model storage unit 31 stores human body model information. The equipment model storage unit 32 stores equipment model information. The human body model information is the information used in arranging the human body model of a standard (or reference) physique in the three-dimensional virtual space. By reading the human body model information, the simulation unit 19 arranges the human body model of the standard physique in the three-dimensional virtual space. The equipment model information is the information used in arranging the equipment model to be evaluated in the three-dimensional virtual space. By reading the equipment model information, the simulation unit 19 arranges the equipment model in the three-dimensional virtual space. The view determination unit 20 performs a view determining process for determining whether or not a visual identification target part of the equipment model displayed on the view image is in the range of the view of the human body model, that is, whether or not the human body model can visually confirm the visual identification target part. That is, the view determination unit 20 is a view determination unit determining whether or not a visual identification target part is in the range of the view of the human body model using visual identification management unit managing as visual identification information the information about the visual identification target part as a part of the equipment model to be visually confirmed by the human body model. The view determination unit 20 includes a visual identification determination unit 201, a character recognition determination unit 202, and a next operation determination unit 203, calls them to perform the visual identification determining process, the character recognition determining process, and the next operation determining process described later. That is, the view determination unit 20 is visibility/readability determination unit (a first determination unit) determining the visibility of the visual identification target part of the equipment model displayed on the view image and the readability of the character string associated with the visual identification target part based on the information about the visual identification target part. The operation determination unit 21 determines the operability of the equipment in a series of operations performed by the human body model based on the determination result performed by the view determination unit 20. That is, the operation determination unit 21 is an operability determination unit (a second determination unit) determining the operability of the equipment based on the determination result by the view determination unit 20. The operation determination unit 21 includes the operation determination table 30, and records the determination result by the view determination unit 20 and the determination result of the operability of the equipment in the operation determination table 30 for each operating posture of the human body model. Therefore, the operation determination table 30 stores the determination result by the operation determination unit 21. In the view determination unit 20, the visual identification determination unit 201 performs a visual identification determining process for determining whether or not a visual identification target part on the view image can be visually identified (for example, easily seen). The result of the visual identification determining process by the visual identification determination unit 201 is recorded in the visual identification management table 29 provided for the view determination unit 20. The character recognition determination unit 202 performs a character recognition determining process for determining the readability of a character string displayed on the visual identification target part in the view image. The character recognition determination result is recorded in the character string setting table 27 provided for the character string setting unit 16. The next operation determination unit 203 performs a next operation determining process of determining whether or not the range of the view of the human body model in the view image includes the target position or the target part of the next operation. The equipment operability evaluation apparatus 1 is formed by a computer, and the function of each component is realized by a CPU and a program executed by the CPU. The program can be stored in a computer-readable recording medium, for example, semiconductor memory, a hard disk, CD-ROM, a DVD and so on, can be recorded in any of these recording media and provided, or provided by the transmission and reception using a network through a communication interface. FIG. 2 shows an example of an explanatory view of the inverse kinematics (IK). In FIG. 2, among human body parts of the human body model 100, a portion to be moved to the point P as a target position (hereinafter referred to as an end-effector) 300 corresponds to a control point. Reference numeral 301 designates a fixed portion, and θ1 indicates the angle of the shoulder joint of the human body model 100. The calculation range of the inverse kinematics is the range from the end-effector 300 to the last fixed portion 301 through each parent link. That is, the inverse kinematics calculates a posture (subsequent posture) from a certain posture (hereinafter referred to as a preceding posture) of the human body model 100, and the calculation result is an angle of each joint of the human body model 100. The calculation of a posture is performed using the position of the fixed point (omitted in the drawings) of the fixed portion 301, the angle of each joint of the shoulder joint, the elbow, the wrist, the position of the end-effector 300, and the preceding posture. When the posture to be calculated is the human body model 100, normally the waist is set as a fixed point because it is positioned at the center of the limbs. Assume that the human body model 100 takes the posture for allowing the end-effector 300 to move to the point P as a target position. When the size of the human body model 100 is changed, the position of the waist also changes. If the fixed point of the human body model 100 before the change of the size is set to the position of the waist, the position of the end-effector 300 is separated from the point P. Using the inverse kinematics, the posture for allowing the end-effector 300 to move to the point P as a target position can be calculated for any human body type by associating the fixed point with the target position. FIG. 3 shows an example of a data structure of a physique management table. For each “human body type”, the physique management table 23 stores “height”, “brachium”, “forearm”, “active”, “human body type ID” and so on as associated with the “human body type”. The “human body type” stores the type of the statistical value (for example, the value of a height, the value of the length of a brachium, the value of the length of a forearm and so on) of each part of a human body stored in the physique management table 23. For example, as shown in FIG. 3, the “human body type” stores a “Japanese female 5% tile value” indicating the 5% tile value of a Japanese female, and a “North American male 99% tile value” indicating the 99% tile value of a North American male. The “height” stores the value of the height corresponding to the human body type stored in the “human body type”. The “brachium” stores the value of the length of a brachium corresponding to the human body type stored in the “human body type”. The “forearm” stores the value of the length of a forearm corresponding to the human body type stored in the “human body type”. The “active” sets a flag indicating which human body type has been selected. For example, when a human body type is selected, a flag indicating that the human body type has been selected is set in the “active” corresponding to the selected human body type is set. In the example in the physique management table 23 shown in FIG. 3, the Japanese male 50% tile value is selected as a human body type. The “human body type ID” stores an identifier uniquely identifying a human body type. FIG. 4 shows an example of the data structure of a target management table. The target management table 24 stores “item”, “control point”, “target part” and so on as associated with the “target ID” for each “target ID”. The “target ID” stores an identifier uniquely identifying the target information. The “item” stores a number of each record of the target management table 24. The number of the record is a number of the row in the target management table 24. A record refers to each row (one line) in the target management table 24 (same in other tables). The “control point” stores a name of the control point as a part of a human body model to be moved to a target position. The “target position” stores coordinates of the target position. The coordinates of the stored target position are, for example, the relative coordinates from a reference position of the human body model. In the example of the target management table 24 shown in FIG. 4, for example, the target ID of T1 indicates the target information that the right hand of a human body model is to be moved to the target position of the coordinates (100, 200, 1000). FIG. 5 shows an example of the data structure of the visual identification management table. The visual identification management table 26 stores an “item”, a “visual identification target part 1”, a “visual identification target part 2” and so on as associated with the “visual identification ID” for each “visual identification ID”. The “visual identification ID” stores an identifier uniquely identifying the visual identification information. The “item” stores a number of each record in the visual identification management table 26. The “visual identification target part 1” and the “visual identification target part 2” store a name of a visual identification target part as a part to be visually confirmed by a human body model. FIG. 6 shows an example of the data structure of the target part management table. The target part management table 25 stores an “item”, a “target part”, a “target ID”, a “relative position from target position” and so on as associated with a “target part ID” for each “target part ID”. The “target part ID” stores an identifier for uniquely identifying a target part information. The “item” stores a number of each record of the target part management table 25. The “target part” stores a name of a target part. The “target ID” stores a target ID corresponding to the target position to be associated with the target part. The “relative position from target position” stores relative coordinates of the target part stored in the “target part” from the target position corresponding to the target ID stored in the “target ID”. By setting the “relative position from target position”, the target part stored in the “target part” is associated with the target position indicated by the target ID stored in the “target ID”. For example, as indicated by the first record of the target part management table 25 shown in FIG. 6, when the “target part” is C. part, and the “target ID” is T1, “(10, 20, 30)” stored in the “relative position from target position” indicate the relative coordinates of the C. part from the target position “(100, 200, 1000)” corresponding to the target ID of T1 in the target management table 24. FIG. 7 shows an example of a data structure of the character string setting table. The character string setting table 27 stores the “item”, a “position X, Y”, a “size”, a “character string length”, a “row spacing”, a “character line width”, a “character color RGB”, a “determination” and so on as associated with the “visual identification ID” for each “visual identification ID”. The “visual identification ID” stores a visual identification ID corresponding to the visual identification target part on which a character string as a target of a character recognition determination is displayed. The “item” stores a number of each record of the character string setting table 27. The “position X, Y” stores coordinates on the view image of the character string displayed on the visual identification target part corresponding to the “visual identification ID”. The “size” stores a size of a character string (for example, a point number and so on). The “character string length” stores a length of a character string. The “row spacing” stores a row spacing of a character string. The “character line width” stores a line width of a character forming the character string. The “character color RGB” stores a color of a character of a character string (for example, an RGB value). The “determination” stores a result of the character recognition determining process by the character recognition determination unit 202 for the character string. An RGB value refers to the value of a gray scale (for example, 0 to 255) of each signal of R (red), G (green), and B (blue) when the color is expressed by the three primary colors (RGB). FIG. 8 shows an example of a data structure of the operation management table. The operation management table 28 stores the “item”, the “target ID”, a “visual identification ID 1”, a “visual identification ID 2”, a “visual identification ID 3” and so on as associated with the “order” for each “order”. Each record of the operation management table 28 corresponds to each operating posture of the human body model performing a series of operations. The “order” stores an order of each operating posture taken by the human body model. The “item” stores a number of each record of the operation management table 28. The “target ID” stores a target ID corresponding to each operating posture. The “visual identification ID 1” to “visual identification ID 3” store the visual identification ID corresponding to each operating posture. For example, in the first record of the operation management table 28 shown in FIG. 8, T1 is set for the “target ID”, S1 is set for the “visual identification ID 1”, and S2 is set for the “visual identification ID 2”. On the target management table 24 shown in FIG. 4, the control point corresponding to the target ID of T1 refers to the right hand, and the target position is indicated by the coordinates (100, 200, 1000). Referring the visual identification management table 26 shown in FIG. 5, the visual identification target parts corresponding to the visual identification ID of S1 are A. part and C. part, and the visual identification target part corresponding to the visual identification ID of S2 is B. part. Therefore, for example, the first record of the operation management table 28 shown in FIG. 8 corresponds to the operating posture of the human body model moving its right hand and the watch point to the coordinates (100, 200, 1000) and visually confirming the A. part, C. part, and B part. FIG. 9 shows an example of a data structure of the visual identification determination management table. The visual identification management table 29 stores the “item”, the “visual identification part color RGB”, the “extracted background color RGB”, the “brightness difference”, the “color difference”, the “presence/absence of character string”, the “determination” and so on as associated with the “visual identification ID” for each “visual identification ID”. The “visual identification ID” stores a visual identification ID corresponding to the visual identification target part as an object of the visual identification determining process by the visual identification determination unit 201. The “item” stores a number of each record of the visual identification management table 29. The “visual identification part color RGB” stores color information (for example, an RGB value) about the visual identification target part corresponding to the “visual identification ID”. The “extracted background color RGB” stores color information (for example, an RGB value) of the background area in the view image of the visual identification target part corresponding to the “visual identification ID”. The “brightness difference” stores information indicating whether or not the brightness difference between the visual identification target part and the background area of the visual identification target part is equal to or exceeds a predetermined threshold. For example, the “brightness difference”-stores information indicating whether or not the brightness difference is equal to or exceeds 125. The “color difference” stores information indicating whether or not the color difference between the visual identification target part and the background area of the visual identification target part is equal to or exceeds a predetermined threshold. For example, the “color difference” stores the information indicating whether or not the color difference is equal to or exceeds 500. The “presence/absence of character string” stores information indicating whether or not a character string is displayed on the visual identification target part corresponding to the “visual identification ID”. The “determination” stores a result of the visual identification determining process by the visual identification determination unit 201 for the visual identification target part corresponding to the “visual identification ID”. FIG. 10 shows an example of a data structure of the operation determination table. The operation determination table 30 stores the “item”, the “order”, the “view target”, the “view determination result”, the “visual identification determination result”, the “character recognition determination result”, the “next operation determination result”, the “operability determination result” and so on as associated with the “human body type” for each “human body type”. The “item” stores a number of each record of the operation determination table 30. The “human body type” stores a human body type ID corresponding to the human body type of the human body model operating the equipment model. The “order” stores an order in which the human body model takes the operating posture. The “view target” stores a visual identification ID of the visual identification target part as an object of the view determining process, the visual identification determining process, and the character recognition determining process. That is, the “view target” stores a visual identification ID corresponding to the operating posture associated with the order stored in the “order”. For example, the “view target” corresponding to the first order stores the visual identification IDs of S1 and S2 stored in the first record of the operation management table 28 shown in FIG. 8. The “view determination result” stores a result of the view determining process by the view determination unit 20. The “visual identification determination result” stores a result of the visual identification determining process by the visual identification determination unit 201. The “character recognition determination result” stores a result of the character recognition determining process by the character recognition determination unit 202. The “next operation determination result” stores a result of the next operation determining process by the next operation determination unit 203 as associated with the “order”. The “operability determination result” stores a determination result of the operability of the equipment of the operation determination unit 21 as associated with the “order”. The change in the operating posture of the human body model when the human body type is changed is explained below by referring to FIG. 11. FIG. 11A shows an example of the operating posture taken by the human body model before the change of the human body type. FIG. 11B shows an example of the operating posture taken by the human body model when the human body type is changed. In FIGS. 11A and 11B, the point P denotes the target position of the human body model 100 set in the target management table 24, 302 denotes a target part set in the target part management table 25, and 303 to 305 denote visual identification target parts 303 to 305 are set in the visual identification management table 26. The target part management unit 14 keeps the relative position for the point P as a target position of the target part 302 constant. For example, the posture calculation unit 18 calculates the posture for allowing the right hand as a control point and the watch point of the human body model 100 to move from the target part 302 to the point P in a predetermined relative position based on the physique information corresponding to the human body type selected from the physique management table 23, the target information set in the target management table 24, and the target part information set in the target part management table 25. In one embodiment of the present invention, the posture calculation unit 18 can also calculate the posture for allowing the right hand as a control point and the watch point of the human body model 100 to move from the target part 302 to the point P in a predetermined relative position and for allowing the human body model 100 to virtually confirm the visual identification target parts 303 to 305 based on the physique information corresponding to the human body type selected from the physique management table 23, the target information set in the target management table 24, the target part information set in the target part management table 25, and the visual identification information set in the visual identification management table 26. Based on the calculated posture information, the simulation unit 19 allows the human body model 100 to take the posture in the three-dimensional virtual space as shown in FIG. 11A. When the human body type of the human body model 100 is changed, the simulation unit 19 changes the physique of the human body model 100 based on the physique information corresponding to the changed human body type in the physique management table 23. The posture calculation unit 18 calculates the posture for allowing the right hand as the control point of the human body model 100 having the changed physique to move to the point P. The simulation unit 19 allows the human body model 100 having the changed physique to take the posture shown in FIG. 11B in a three-dimensional virtual space. FIG. 12 shows the operating posture of a human body model when the layout of an equipment model is changed. For example, assume the case in which the layout of the equipment model 101 shown in FIG. 11A is changed, and the position of the target part 302 as an operation target is moved to the position below the position of the target part 302 shown in FIG. 11A as shown in the layout of the equipment model 101 in FIG. 12. The posture calculation unit 18 changes the coordinates of the target position associated with the target part 302 so that the relative position to the target position of the target part 302 set in the target part management table 25 can be kept constant. For example, the coordinates of the point P as a target position in FIG. 11A are changed to the coordinates of the point R shown in FIG. 12. The posture calculation unit 18 calculates the posture for allowing the right hand as a control point and the watch point of the human body model 100 to move to the point R at a predetermined relative position from the target part 302 whose position is changed. In one embodiment of the present invention, the posture calculation unit 18 can calculate the posture for allowing the right hand and the watch point of the human body model 100 as a control point to move to the point R at a predetermined relative position from the target part 302 whose position is changed and for allowing the human body model 100 to virtually confirm the visual identification target parts 303 to 305. Based on the calculated posture information, the simulation unit 19 allows the human body model 100 to take the posture shown in FIG. 12 in a three-dimensional virtual space. FIGS. 13 and 14 show an example of a flowchart of an equipment operability evaluating process performed by the equipment operability evaluation apparatus 1 according to the present invention, and collectively show the flowchart of the process. In FIG. 13, the simulation unit 19 calls from the equipment model storage unit 32 an equipment model whose operability is to be evaluated according to the specification input of the user of the equipment operability evaluation apparatus 1 (step S1). For example, the simulation unit 19 calls a model of an ATM. Then, the simulation unit 19 selects any human body type from the physique management table 23, extracts from the human body model storage unit 31 a human body model corresponding to the physique of the selected human body type, and arranges the extracted human body model together with the equipment model called in the step S1 in the three-dimensional virtual space (step S2). Next, the target management unit 13 sets a control point and a target position in the target management table 24 according to the specification input of the user (step S3). The target management unit 13 (or the target part management unit 14) checks whether or not there is a target part relating to the set target position (step S4). When there is a target part, the target part management unit 14 sets the target part as associated with the target position in the target part management table 25 according to the specification input of the user (step S5). For example, the name of the target part and the relative position from the target position are set as associated with the target ID in the target part management table 25. When there is no target part relating to the target position, the process in the step S5 is omitted. Next, the target management unit 13 (or visual identification management unit 15) checks whether or not there is a visual identification target part (step S6). When there is a visual identification target part, the visual identification management unit 15 sets a visual identification target part in the visual identification management table 26 according to the specification input of the user (step S7). Then, the target management unit 13 (or character string setting unit 16) checks whether or not there is (included) a character string on the visual identification target part set in step S7 (step S8). When there is a character string, the character string setting unit 16 sets the attribute of a character string in the character string setting table 27 according to the specification input of the user (step S9). When there is no character string in the visual identification target part, the process in the step S9 is omitted. When there is no visual identification part in the step S6, the processes in the steps S7 to S9 are omitted. Then, the operation management unit 17 sets the operation information in the operation management table 28 according to the specification input of the user (step S10). For example, the operation management unit 17 sets the order in which the human body model takes the operating posture, the target ID, and (a plurality of) visual identification IDs in the operation management table 28 for each of the operating posture taken by the human body model. Afterwards, the target management unit 13 checks whether or not there is a next operating posture to be taken by the human body model (step S11). When there is a next operating posture, the processes in and after the step S3 are repeated. When there is no subsequent operation (when the operation information has been set for all operating postures), the process in step S12 shown in FIG. 14 is repeated. In FIG. 14, the posture calculation unit 18 calculates the posture to be taken by the human body model based on the operation information stored in the operation management table 28 (step S12). For example, the target ID of T1 and the visual identification IDs of S1 and S2 are set in the operation management table 28 shown in FIG. 8 corresponding to the operating posture to be firstly taken by the human body model. Furthermore, in the target management table 24 shown in FIG. 4, the control point as the right hand and the target position having the coordinates (100, 200, 1000) are set corresponding to the target ID of T1. Then, for example, the posture calculation unit 18 calculates the posture of the human body model for allowing the right hand of the human body model to move to the coordinates (100, 200, 1000). Next, the simulation unit 19 allows the human body model to take an operating posture based on the posture information calculated by the posture calculation unit 18, and displays the view image as an image of a three-dimensional virtual space viewed from the point of view of the human body model (step S13). In the process in the step S13, for example, the view image shown in FIG. 15 is displayed. In FIG. 15, the point A is a target position for the current operating posture of the human body model, the point B is the target position for the next operating posture, a flicker 103 and labels 104 and 105 are the visual identification target parts of the equipment model 101 for the current operating posture. In the visual identification management table 26 shown in FIG. 5, the visual identification target part of A. part and C. part are set corresponding to the visual identification ID of S1, and the visual identification target part of B. part is set corresponding to the visual identification ID of S2. Therefore, for example, A. part, B. part, and C. part respectively correspond to the flicker 103, label 104, and label 105 as visual identification target parts as shown in FIG. 15. Additionally, when the current operating posture is the first operating posture taken by the human body model, then the target ID of T3 corresponding to a second operating posture set in the operation management table 28 shown in FIG. 8 indicates the coordinates of the point B. The oval shown in the view image shown in FIG. 15 is the range of the view of the human body model. Next, the view determination unit 20 performs the view determining process (step S14). For example, the view determination unit 20 determines whether or not the flicker 103 and the labels 104 and 105 shown in FIG. 15 are in the range of the view of a human body model. Then, the visual identification determination unit 201 performs the visual identification determining process (step S15). For example, the visual identification determination unit 201 determines in the following process as to whether or not the flicker 103 as the visual identification target part shown in FIG. 15 can be visually identified (easily seen). That is, the visual identification determination unit 201 extracts the contour of the flicker 103 from the view image shown in FIG. 15, and generates a banding box as shown in FIG. 16A. Then, the visual identification determination unit 201 generates an enlarged image 1.1 time larger than the banding box shown in FIG. 16A with any point of the flicker 103 in the banding box as the center. FIG. 16B shows the generated enlarged image. The visual identification determination unit 201 superposes the banding box shown in FIG. 16A on the enlarged image shown in FIG. 16B as shown in FIG. 16C. The diagonal portion shown in FIG. 16C is an area in which the enlarged image and the banding box do not overlap each other. The visual identification determination unit 201 compares, for example, the color (color of the background area of the flicker 103) of the pixel in the view image shown in FIG. 15 for the area of the diagonal portion shown in FIG. 16C with the color (that is, the color of the flicker 103 as a visual identification target part) of the pixel in the enlarged image. Furthermore, the visual identification determination unit 201 compares, for example, the brightness (brightness in the background area) of the pixel in the view image shown in FIG. 15 for the area of the diagonal portion shown in FIG. 16C with the brightness (that is, the brightness of the flicker 103 as a visual identification target part) of the pixel in the enlarged image. The visual identification determination unit 201 firstly extracts the color of the flicker 103 and the color of the background area, and stores the RGB value of the extracted color of the flicker 103 and the RGB value of the extracted color of the background area respectively in the “visual identification part color RGB” and the “extracted background color RGB” in the visual identification management table 29 shown in FIG. 9. The visual identification determination unit 201 compares the RGB value of the color of the flicker 103 with the RGB value of the color of the background area, sets larger R, G, and B values as maxR, maxG, and maxB, respectively, and smaller R, G, and B values as minR, minG, and minB, respectively, and calculates the color difference X1 by using the following equation.X1=(maxR−minR)+(maxG−minG)+(maxB−minB) Additionally, the visual identification determination unit 201 defines the value of the difference between the value obtained by the following equation using the RGB value of the extracted color of the flicker 103 and the value obtained by the following equation using the RGB value of the extracted color of the background area as a brightness difference X2.((R×299)+(G×587)+(B×114))/1000 For example, the visual identification determination unit 201 determines that a visual identification target part is easily seen when the calculated color difference X1 is equal to or exceeding a predetermined threshold (for example, 500) and the calculated brightness difference X2 is equal to or exceeding a predetermined threshold (for example, 125), and decides that the result of the visual identification determining process is OK (W3C Techniques For Accessibility Evaluation And Repair Tools. Refer to Checkpoint 2.2). The visual identification determination unit 201 decides that the visual identification determining process is NG when the calculated color difference X1 is smaller than the predetermined threshold (for example, 500) or the calculated brightness difference X2 is smaller than the predetermined threshold (for example, 125). The visual identification determination unit 201 can also perform the visual identification determining process on a visual identification target part based on the comparison result between the color difference X1 and the predetermined threshold or the comparison result between the brightness difference X2 and the predetermined threshold. The visual identification determination unit 201 stores the result of the visual identification determining process in the visual identification management table 29. In the step S15, the visual identification determination unit 201 can also store in the visual identification management table 29 the information indicating whether or not there is a character string on the visual identification target part. Next, the character recognition determination unit 202 performs a character recognition determining process (step S16). For example, the character recognition determination unit 202 determines in the following process whether or not the character string “card” displayed on the label 104 as a visual identification target part shown in FIG. 15 can be read. That is, the character recognition determination unit 202 extracts the image of the label 104 on which the character string “card” is displayed as shown in FIG. 17A from the data of the equipment model currently arranged in the three-dimensional virtual space stored in advance in the equipment model storage unit 32. Then, the character recognition determination unit 202 generates a substitute image as a virtual image replacing the image of the character string “card” according to the attribute information about the character string “card” of the label 104 set in the character string setting table 27. The image of the rectangular contour line shown in FIG. 17B is a substitute image. The vertical length of the rectangle forming a replacing image is the length corresponding to the size of the character string “card” set in the character string setting table 27, the longitudinal length of the rectangle is the length corresponding to the length of the character string “card”, the width of the contour line of the rectangle corresponds to the width of the character line of the character string “card”, and the color of the contour line of the rectangle is the same as the character color of the character string “card”. The character recognition determination unit 202 generates an image in which the color of the background of the substitute image is the ground color of the equipment model 101 shown in FIG. 15 as shown in FIG. 17C based on the image shown in FIG. 17B. Next, the character recognition determination unit 202 generates an image obtained by viewing the image shown in FIG. 17C in the same direction as the line of sight of the human body model seeing the label 104 shown in current FIG. 15 as shown in FIG. 17D in a three-dimensional virtual space. The substitute image shown in FIG. 17D is distorted depending on the direction of the line of sight of the human body model. The character recognition determination unit 202 acquires an image when light is emitted from the light source set in advance to the image shown in FIG. 17D as shown in FIG. 17E. The image is generated by the simulation unit 19. The substitute image shown in FIG. 17E cannot be seen in the right area by the influence of the emitted light. The character recognition determination unit 202 compares the substitute image shown in FIG. 17D with the substitute image shown in FIG. 17E, and calculates the range in which the character string “card” displayed on the label 104 can be read. The range in which the character string “card” can be read is, for example, the shaded area in the image shown in FIG. 17F. When the calculated range in which the character string “card” can be read is equal to or smaller than a predetermined threshold, the character recognition determination unit 202 decides that the result of the character recognition determining process on the character string “card” is NG. When the calculated range in which the character string “card” can be read is larger than the predetermined threshold, the character recognition determination unit 202 decides that the result of the character recognition determining process on the character string “card” is OK. According to the above-mentioned character recognition determining process, the readability of the character string displayed on the visual identification target part when the visual identification target part of equipment whose operability is to be evaluated is illuminated can be determined. By changing the direction in which light is emitted to the substitute image shown in FIG. 17(D), and performing the character recognition determining process on the character string “card”, for example, it can be determined in which direction the light is emitted to the visual identification target part of the equipment with poor readability of the character string displayed on the visual identification target part. The character recognition determination unit 202 can be configured to determine the readability of the character string “card” based on the actual distance M from the eye of the human body model arranged in the three-dimensional virtual space to the visual identification target part, the actual line width X of the contour of a substitute image, a predetermined eyesight of the human body model, and the direction θ of the line of sight when the human body model sees the visual identification target part. For example, the character recognition determination unit 202 calculates the distance H from the human body model to the substitute image of the character string “card” when the contour of the substitute image of the line width X×sin θ cannot be seen from the human body model having the eyesight of A. X×sin θ is the line width of the contour of the substitute image seen from the human body model taking an operating posture. For example, the line width corresponds to the line width of the contour of the substitute image shown in FIG. 17D. In calculating the distance H, for example, the principle of the vision test using the Landolt ring. Using the principle of the vision test by the Landolt ring, for example, the distance H is calculated by the equation H=2(X×sin θ)×A/(2π/360/60). Then, the character recognition determination unit 202 determines that a substitute image cannot be easily seen when the value of the distance M is larger than the value of the distance H, and decides that the result of the character recognition determining process on the label 104 on which character string “card” is displayed is NG. When the value of the distance M is equal to or smaller than the value of the distance H, the character recognition determination unit 202 decides that the result of the character recognition determining process on the label 104 on which the character string “card” is displayed is OK. According to the above-mentioned character recognition determining process, for example, when the result of the character recognition determining process on the label 104 on which the character string “card” is displayed is NG, it is determined that the character string cannot be easily read in the size of the character string “card” set in the character string setting table 27. That is, according to the above-mentioned character recognition determining process, it can be verified for each operating posture of the human body model whether or not the character string displayed on the visual identification target part of the equipment whose operability is to be evaluated has the size in which the character string can be easily read from the view point of the human model. Next, the target management unit 13 (or the posture calculation unit 18, or the view determination unit 20) checks whether or not there is the next operating posture to be taken by the human body model (step S17). When there is the next operating posture, the next operation determination unit 203 performs the next operation determining process (step S18), and the processes in and after the step S12 are repeated. In the step S18, the next operation determination unit 203 refers to the operation management table 28, identifies the target ID corresponding to the next operating posture to be taken by the human body model, refers to the target management table 24, and identifies the target position corresponding to the target ID. When the identified target position is in the range of the current view of the human body model, the next operation determination unit 203 decides that the result of the next operation determining process is OK. For example, on the view image shown in FIG. 15, the point B as the next target position is in the current range of the view indicated by an oval. Therefore, it is decided that the result of next operation determining process is OK. According to the next operation determining process in the step S18, it can be verified whether or not the display or the sign for indicating the flow of operation of the equipment is provided on the equipment. When there is no next operating posture to be taken by the human body model in the step S17, the operation determination unit 21 determines the operability of the equipment in a series of operations performed by the human body model based on the result of the view determining process, the result of the visual identification determining process, the result of the character recognition determining process, and the result of the next operation determining process (step S19). The operation determination unit 21 records on the operation determination table 30 the result of the view determining process, the result of the visual identification determining process, the result of the character recognition determining process, the result of the next operation determining process, and the result of the determination on the operability of the equipment for each operating posture of the human body model. Afterwards, the target management unit 13 (or the posture calculation unit 18, or the view determination unit 20) checks whether or not there is a human body type not selected from the physique management table 23 (step S20). When there is a human body type not yet selected (the process has not been completed on all human body types), the posture calculation unit 18 selects other human body types in the physique management table 23 (step S21), and the processes in and after the step S12 are repeated. When there is no human body type not yet selected (the process has been completed on all human body types), the output unit 22 outputs the result of the determination on the operability of the equipment (step S22). For example, the output unit 22 outputs the contents of the record in the operation determination table 30. As described above, according to this example, after a human body model is allowed to sequentially take the postures of performing a series of operations by inserting a card, selecting the contents of a transaction, inserting a passbook, inputting the amount of money, and inputting the password shown in FIG. 18, the view determining process, the visual identification determining process, the character recognition determining process, and the next operation determining process can be performed in each operating posture. Particularly, in one embodiment of the present invention, the order in which the operating postures taken by the human body model can be changed only by changing the order in which the human body model takes an operating posture set in the operation management table 28, and the above-mentioned view determining process, the visual identification determining process, the character recognition determining process, and the next operation determining process can be automatically performed for each operating posture taken by the human body model in the changed order. As described above, according to the apparatus and method for evaluating equipment operability of the present invention, the visibility and the readability can be determined on a visual identification target part of an equipment model viewed from the point of view of a human body model by allowing the human body model to automatically take an operating posture adapted to a target position although the physique of the human body model is changed. Additionally, according to the apparatus for evaluating equipment operability of the present invention, when the position of a target part is moved, the target position is changed such that a relative position to the predetermined target part can be kept constant when the position of the target part moves. Therefore, although the layout of equipment is changed and the position of the target part is moved, the human body model can be allowed to automatically take an operating posture adapted to the change of the layout of the equipment. Furthermore, according to the apparatus for evaluating equipment operability of the present invention, the visibility and the readability on a visual identification target part on the equipment model in a series of operations required to attain the purpose of the operation of the equipment can be determined. |
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063273241 | description | DESCRIPTION OF THE PREFERRED EMBODIMENTS FIG. 1 shows a fuel assembly according to the invention. During operation, the fuel assembly is arranged vertically in the reactor core. The fuel assembly comprises an upper handle 1, a lower end portion 2 and a plurality of fuel units 3a, 3b, 3c and 3d stacked one above the other. The fuel unit comprise a plurality of fuel rods 4a, 4band 4c arranged between a top tie plate 5 and a bottom tie plate 6. The fuel units are stacked on top of each other in the longitudinal direction of the fuel assembly and they are stacked in such a way that the top tie plate 5 in one fuel unit is facing the bottom tie plate 6 in the next fuel unit in the stack. A fuel rod contains fuel in the form of a stack of uranium pellets 8 arranged in a cladding tube 7. The fuel assembly is enclosed in a fuel channel 9 with a substantially square cross section. In this embodiment, the fuel assembly comprises four parallel stacks with ten fuel units in each stack. FIG. 2a shows a section B--B through the fuel assembly in FIG. 1. The fuel channel 9 is provided with a hollow support member 10 of cruciform cross section, which is secured to the four walls of the fuel channel. In the central channel 11 formed of the support member 10, moderator water flows. The fuel channel with the support member surrounds four vertical channel-formed parts 12a, 12b, 12c, 12d, so-called sub-channels, with an at least substantially square cross section. The four sub-channels each comprises a stack of fuel units. Each fuel unit comprises 24 fuel rods arranged in a symmetrical 5.times.5 lattice. By a fuel rod position is meant a position in the lattice. All the fuel rod positions in the lattice need not be occupied by fuel rods. The fuel assembly has three different types of fuel rods 4a, 4b and 4c. In FIGS. 2a-2d the fuel rods 4a are designated M and the fuel rods 4a are designated P. The fuel rods 4b are not marked in the figures. A fuel rod 4a has a diameter d.sub.1. A fuel rod 4b has a diameter d.sub.2 which is about 8% larger than d.sub.1 and contains about 15% more fuel than the fuel rod 4a. A fuel rod 4c has a diameter d.sub.3 which is about 8% larger than d.sub.2 and contains about 15% more fuel than the fuel rod 4b. By varying between the three fuel rod types in the different lattice positions, a great variation of fuel units may be created. The fuel rods with the largest diameter, 4c, have a relatively larger fission gas space than the fuel rods with the smallest diameter, 4a, in order thus to take into account different linear loads because of rod diameters and typical neutron-flux ratios. It is not sufficient that the diameter is larger but also the height of the fission gas space should be larger. In this embodiment, the fuel assembly is composed of four different types of fuel units 3a, 3b, 3c, 3d at ten different levels. FIG. 2d shows a section E--E through the fuel unit 3a. The fuel unit 3d is formed to fit into the lower part of the fuel assembly where the neutron flux tends to be high for a large part of the operating cycle. This fuel unit almost exclusively comprises fuel rods of the 4c type, which is that of that fuel rods which has the largest cross-section area and contains most fuel. In the lowermost part of the fuel assembly, the significance of a reduced flow area because of the large cross-section area of the fuel rods is not so great since both the moderation and the cooling are good and the pressure drop is still low because of a low steam content. The higher up in the fuel assembly, the fewer are the fuel rods with the largest diameter 4c and instead the number of fuel rods with a smaller diameter 4a and 4b increases. FIG. 2b shows a section C--C through the fuel unit 3c and FIG. 2c shows a section D--D through the fuel unit 3bFIG. 2a shows a section B--B through the uppermost fuel unit 3d, which comprises only fuel rods of types 4a and 4b, which both have a diameter and a fuel content which are smaller than those of the fuel rod 4c. In addition, the lattice positions nearest the water channel 11 are unoccupied. One advantage of the unoccupied positions is that the shutdown margin increases. In the upper part of the fuel assembly, the optimization of the fuel units takes place in order to minimize the risk of dryout and to obtain a low pressure drop. To absorb part of the surplus reactivity in the fuel when it is fresh, certain of the fuel rods may contain a burnable absorber, for example gadolinium oxide. Such a fuel rod will be referred to below as an absorber rod. The diameter of the absorber rod determines its burnup rate. The absorber rods 13a, 13b, 13c are available in three different sizes with three different diameters d.sub.1, d.sub.2, d.sub.3 which are the same as for the fuel rods. By arranging absorber rods with different diameters in the lattice, the content of burnable absorber may be finely-divided both axially and laterally with respect to reactivity, burnup behavior and power distribution. FIG. 3 shows an absorber rod 13c in cross section. The absorber rod comprises a plurality of fuel pellets 15 and 8a stacked on top of each other in a cladding tube 7 and a top plug 16 and a bottom plug 17 seal the absorber rod. The fuel. pellets 15 contain a certain part of a burnable absorber. The two end pellets 8a in the absorber rod only contain fuel and lack burnable absorber. The end pellets in both the fuel rods and the absorber rods adjoin axial gaps which arise between the fuel units in the stack. Because of the axial gap, the moderation and hence the reactivity become higher in the end pellets compared with the other pellets in the stack. By not adding any burnable absorber to the fuel in the end pellets, the end pellets in the fuel unit are burnt up faster than other pellets. The burnup takes place at the beginning of the operating cycle while the total power of the fuel assembly is still limited by the burnable absorber. Since it is necessary in some way, for example by a lower enrichment or by providing them with holes, to limit the power in the end pellets of the fuel rods, it is an advantage that all the end pellets are identical so that the manufacture is simplified. In this embodiment, all the fuel units have the same kind of lattice. It is an advantage that all the fuel units have the same lattice because then the same bottom tie plates and top tie plates may be used for the different fuel units, which minimizes the number of components which need to be manufactured and kept in stock. It is also possible, while maintaining the same lattice, to carry out optimizations by limited displacements of the positions of the rods. In another embodiment, fuel units at the same level in the fuel assembly may have different distribution of fuel rods. This may, for example, be advantageous in a reactor where the fuel assembly is surrounded by water gaps with different widths. The moderation becomes different depending on which gaps a fuel unit is facing, which may be compensated for by arranging fuel rods with larger or smaller diameters in lattice positions adjacent the gaps. |
claims | 1. A fuel assembly configured to be positioned in a nuclear water reactor, wherein the fuel assembly comprises:an upstream end,a downstream end,a flow interspace between the upstream end and the downstream end,a plurality of fuel rods provided in the flow interspace between the upstream end and the downstream end, the flow interspace being configured to permit a flow of coolant through the fuel assembly along a flow direction from the upstream end to the downstream end in contact with the fuel rods,a filter device configured to catch debris particles in the flow of coolant,wherein the filter device comprises a first filter zone provided in the flow of coolant between the upstream end and the fuel rods, and a second filter zone,wherein the first filter zone comprises a plurality of passages arranged to guide at least a part of the flow of coolant to pass the first filter zone through the passages towards the downstream end,wherein the second filter zone comprises a plurality of passages arranged to guide at most a part of the flow of coolant to pass the second filter zone through the passages towards the downstream end,wherein the first filter zone has a first filtering efficiency and the second filter zone has a second filtering efficiency,wherein the second filtering efficiency is higher than the first filtering efficiency, andwherein the passages of the first filter zone and the second filter zone are formed by a plurality of sheets, which are arranged beside each other and oriented along the flow direction, and wherein adjacent sheets of the first filter zone are provided at a larger distance from each other than adjacent sheets of the second filter zone. 2. The fuel assembly according to claim 1, wherein the second filter zone is dimensioned to permit the second filter zone to be clogged and wherein the first filter zone is configured to secure a sufficient flow of coolant through the first filter zone even if no coolant passes the second filter zone. 3. The fuel assembly according to claim 1, wherein the first filter zone comprises an inlet end turned towards the upstream end, and an outlet end, and is arranged to guide said at least a part of the flow of coolant to pass the first filter zone through the passages from the inlet end to the outlet end, andwherein the second filter zone comprises an inlet end turned towards the upstream end, and an outlet end, and is arranged to guide said at most a part of the flow of coolant to pass the second filter zone through the passages from the inlet end to the outlet end. 4. The fuel assembly according to claim 1, wherein the first filter zone and the second filter zone are provided beside each other. 5. The fuel assembly according to claim 1, wherein each of the passages of the first filter zone defines a first flow area and each of the passages of the second filter zone defines a second flow area, which is smaller than the first filter area. 6. The fuel assembly according to claim 1, wherein each of the passages of the first filter zone defines a first passage length from an inlet end to an outlet end, and each of the passages of the second filter zone defines a second passage length from an inlet end to an outlet end, which second passage length is longer than the first passage length. 7. The fuel assembly according to claim 1, wherein the first filter zone has a first pressure loss coefficient ξ1 and the second filter zone a second pressure loss coefficient ξ2 and wherein the second pressure loss coefficient ξ2 is greater than the first pressure loss coefficient ξ1. 8. The fuel assembly according to claim 1, wherein the first filter zone has a first flow area A1 and the second filter zone a second flow area A2 and wherein the first flow area A1 is greater than the second flow area A2. 9. The fuel assembly according to claim 1, wherein the filter device comprises a magnetic member provided to create a magnetic field on at least some of the passages of the second filter zone to attract debris particles flowing through the passages of the second filter zone. 10. The fuel assembly according to claim 1, wherein the filter device is provided in the flow interspace to guide at most a minor part of the flow of coolant through the second filter zone. 11. The fuel assembly according to claim 10, wherein the filter device is provided in the flow interspace to guide at least a major part of the flow of coolant through the first filter zone. 12. The fuel assembly according to claim 1, wherein the filter device is provided in the flow interspace to guide the entire flow of coolant through the first filter zone. 13. The fuel assembly according to claim 12, wherein the second filter zone is provided at a distance from the first filter zone. 14. The fuel assembly according to claim 13, wherein the second filter zone is provided downstream the first filter zone. 15. The fuel assembly according to claim 13, wherein the second filter zone is provided upstream the first filter zone. 16. The fuel assembly according to claim 1, wherein each sheet comprises a first portion, which extends from the inlet end, a second portion, which extends from the outlet end, and a third portion, which extends between the first portion and the second portion, and wherein each sheet along the first portion has a wave-shape extending in a direction transversally to the flow direction and along the third portion has a wave-shape extending in the flow direction. 17. The fuel assembly according to claim 16, wherein each sheet along the second portion has a wave-shape in the direction transversally to the flow direction. |
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summary | ||
claims | 1. A collimator for stray radiation, comprising:a plurality of absorption elements to absorb X-ray radiation, separated from one another by a filler and support material, the plurality of absorption elements being at least one of aligned approximately in parallel and oriented towards a common focus, and the plurality of absorption elements being arranged in a statistically distributed manner, wherein the filler and support material are formed by individual fibers of a material which is transparent to X-ray radiation. 2. The collimator as claimed in claim 1, wherein the absorption elements include individual fibers of a material which is absorbent for X-ray radiation. 3. The collimator as claimed in claim 2, wherein the fibers of the material which is absorbent for X-ray radiation, include a fiber diameter of 0.2 mm. 4. The collimator as claimed in claim 2, wherein the fibers of the material which is absorbent for X-ray radiation and the fibers of the material which is transparent to X-ray radiation are at least one of aligned approximately in parallel with one another and oriented towards the common focus. 5. The collimator as claimed in claim 1, wherein the absorption elements and the filler and support material are bonded to one another. 6. The collimator as claimed in claim 1, wherein the absorption elements and the filler and support material are present in the collimator in a volume ratio which results in a filling percentage of between 5% and 30% with the absorption elements. 7. The collimator as claimed in claim 1, wherein the absorption elements are formed of a metallic material and the filler and support material is a plastic material. 8. The collimator as claimed in claim 1, wherein the collimator is constructed in the form of a level plate in which the absorption elements are aligned at least approximately perpendicularly to a plate plane. 9. The collimator as claimed in claim 1, wherein the collimator is constructed in the form of a plate which is bent approximately in the form of a calotte shell and in which the absorption elements are aligned towards a center of the sphere. 10. A medical X-ray device comprising the collimator of claim 1. 11. A method for producing a collimator including a plurality of absorption elements to absorb X-ray radiation and a material to separate the plurality of absorption elements, the method comprising:bonding the absorption elements to the material to form a collimator in such a manner that a statistical distribution of the absorption elements over a width of the collimator is obtained, wherein fibers of a material which is absorbent for X-ray radiation as absorption elements are intermixed with fibers of a material which is transparent to X-ray radiation as the material, from the intermixed fibers a fiber stack is formed and bonded to form a compound fiber system and the compound fiber system is split into individual discs perpendicularly to the fiber axes of the intermixed fibers. 12. The method as claimed in claim 11, wherein the fibers of the material which is absorbent for X-ray radiation and the fibers of the material transparent to X-ray radiation are intermixed in a ratio which results in a filling percentage of between 5% and 30% with the fibers of the material which is absorbent for X-ray radiation. 13. The method as claimed in claim 11, wherein the absorption elements and the material are bonded to one another. 14. A collimator for stray radiation, comprising:a plurality of absorption elements to absorb X-ray radiation; anda material to separate the plurality of absorption elements, the plurality of absorption elements being arranged in a statistically distributed manner, wherein the material is formed by individual fibers of a material which is transparent to X-ray radiation. 15. A medical X-ray device comprising the collimator of claim 14. 16. The collimator as claimed in claim 14, wherein the absorption elements include individual fibers of a material which is absorbent for X-ray radiation. 17. The collimator as claimed in claim 16, wherein the fibers of the material which is absorbent for X-ray radiation, include a fiber diameter of 0.2 mm. |
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063320118 | abstract | A method of inspecting the H1 weld between the shroud head flange and the upper shroud section utilizing ultrasonic scanning includes the steps of positioning a phased array ultrasonic probe on a top surface of the shroud head flange, emitting an ultrasonic beam from the ultrasonic probe, electronically steering the ultrasonic beam to scan the weld joining the shroud head flange and the upper shroud section with the beam moving from an outer surface of the shroud head flange to an inner surface of the shroud head flange, and acquiring scan data over the length of the scan. The ultrasonic probe is moved circumferentially along the top surface of the shroud head flange in increments of between about 0.05 inch to about 0.5 inch with the H1 weld ultrasonically scanned after each incremental move. |
claims | 1. A method to insert nuclear fuel pellets into a fuel rod cladding element, comprising:providing a cladding element;providing nuclear fuel pellets aligned in a column on a segment make-up table by a pellet pusher device;illuminating a predefined length of the column of nuclear fuel pellets with a laser;removing any non-illuminated nuclear fuel pellet from the segment make-up table;subsequent to the removing step, measuring a length of the column of nuclear fuel pellets with a camera while the column of nuclear fuel pellets is on the segment make-up table, wherein the camera measures the length of the column of nuclear fuel pellets through measuring a distance of an overall position of the pellet pusher device;comparing the length of the column of nuclear fuel pellets to an expected design length; andincorporating the column of nuclear fuel pellets into the cladding element when the measured length of the column of nuclear fuel pellets is within a threshold value of the expected design length. 2. The method to insert nuclear fuel pellets into a fuel rod cladding element according to claim 1, wherein the step of incorporating the column of nuclear fuel pellets into the cladding element is accomplished through vibrating a table upon which the column of nuclear fuel pellets rests such that the vibration causes the column of nuclear fuel pellets to enter into the cladding element. 3. The method to insert nuclear fuel pellets into a fuel rod cladding element according to claim 1, further comprising the step of an operator manually transferring a nuclear pellet sheet from a fuel pellet vault to the segment make-up table, prior to the step of providing nuclear fuel pellets aligned in a column on a segment make-up table by a pellet pusher device. 4. A method to insert nuclear fuel pellets into a fuel rod cladding element, comprising:providing the fuel rod cladding element having a bar code on an exterior of the cladding element;reading the bar code on the cladding element;transporting the cladding element to a rod loader input queue;placing the cladding element on separator rollers, the separator rollers configured to separate the clad from each other;lifting the cladding element onto a vibration table;restraining the clad with a rod holding tool;inserting the cladding element into pellet funnels, the pellet funnels configured to accept fuel pellets and transport the fuel pellets into the clad;providing fuel pellets, the fuel pellets stored in pellet vaults;rotating the pellet vaults to a position to allow an operator to manually remove a pellet sheet containing the fuel pellets;manually removing the pellet sheet from the pellet vault containing the nuclear fuel pellets;deploying a segment stop across a segment make-up table to receive nuclear fuel pellets;discharging nuclear fuel pellets from the pellet sheet onto the segment make-up table, the nuclear fuel pellets positioned against the segment stop;pushing the pellets on the table against the segment stop with a pellet pusher device;illuminating a laser to visually identify which of the nuclear fuel pellets should be incorporated into the cladding element, the laser calibrated to precisely visually indicate an expected length of a segment of nuclear fuel pellets to be incorporated into the cladding element;manually removing nuclear fuel pellets not illuminated by the laser from the table;measuring a cumulative length of fuel pellets in rows remaining on the table through the use of a camera, wherein the camera measures the cumulative length through measuring a distance of an overall position of the pellet pusher device;measuring the cumulative length of the fuel pellets in rows on the table through the use of linear variable differential transformers;verifying the cumulative length of the fuel pellets, as measured by the camera and the linear variable differential transformers, to a design specification of the fuel rod to a correct length;removing fuel pellets from the table which are not verified to the design specification correct length;transferring fuel pellets from the table which have been verified to a vibratory table input queue; andvibratory loading the fuel pellets from the table into the fuel rod cladding. 5. The method according to claim 4, further comprising:releasing the rod holding tool; andchecking a plenum of the clad. 6. The method according to claim 5, wherein the checking of the plenum of the clad includes inserting a calibrated rod into an open end of the fuel rod clad and reading a length of the plenum. 7. The method according to claim 4, further comprising:side lighting the pellet pusher device on the segment make-up table prior to the step of measuring the cumulative length of the fuel pellets in rows remaining on the table through the use of the camera. 8. A method to insert nuclear fuel pellets into a fuel rod cladding element, comprising:providing a cladding element;providing nuclear fuel pellets aligned in a column on a segment make-up table by a pellet pusher device;illuminating a predefined length of the column of nuclear fuel pellets with a laser;removing any non-illuminated nuclear fuel pellet from the segment make-up table;measuring a length of the column of nuclear fuel pellets with at least one of (a) a camera and (b) a linear variable differential transformer while the column of nuclear fuel pellets is on the segment make-up table by measuring a distance of an overall position of the pellet pusher device;comparing the length of the column of nuclear fuel pellets to an expected design length; andincorporating the column of nuclear fuel pellets into the cladding element when the measured length of the column of nuclear fuel pellets is within a threshold value of the expected design length. |
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056132395 | claims | 1. A method for decomposing an organic solution containing radioactive metal ions, which comprises: forming a solution containing radioactive metal ions by applying a solution composed of a chelating solution and/or an organic acid to radioactive metals laden with radioactivity; adding an alkaline agent to the formed solution to elevate electrical conductivity thereof; and electrolyzing the solution elevated in electrical conductivity by an electrolytic device. forming a solution containing radioactive metal ions by applying a solution composed of a chelating solution and/or an organic acid to radioactive metals laden with radioactivity; adding an alkaline agent to the formed solution to elevate electrical conductivity thereof; electrolyzing the solution elevated in electrical conductivity by an electrolytic device to generate therein metal hydroxides which are little soluble in water; and filtering the resultant solution to separate metal hydroxides as a filter cake from the filtrate. forming a solution containing radioactive metal ions by applying a solution composed of a chelating solution and/or an organic acid to radioactive metals laden with radioactivity; adding an alkaline agent to the formed solution to elevate electrical conductivity thereof and to generate therein metal hydroxides which are little soluble in water; filtering the resultant solution to separate the metal hydroxides as a filter cake from the filtrate; electrolyzing the filtrate by an ectrolytic device to further generate therein metal hydroxides which are little soluble in water; and filtering the resultant filtrate to separate the additional metal hydroxides as a filter cake from the filtrate. an electrical conduction tank in which an alkaline agent is fed to a metal solution composed of a chelating solution and/or an organic acid having dissolved therein radioactive metals laden with radioactivity, to elevate electrical conductivity thereof; and an electrolytic device located at a later stage than the electrical conduction tank to electrolyze the metal solution elevated in electrical conductivity. an electrical conduction tank in which an alkaline agent is fed to a metal solution composed of a chelating solution and/or an organic acid containing radioactive metals laden with radioactivity, to elevate electrical conductivity thereof; an electrolytic device located at a later stage than the electrical conduction tank to electrolyze the metal solution elevated in electrical conductivity; and a filtration device positioned at a stage next to and communicating with the electrolyzation device. 2. A method for collecting radioactive metals by way of dissolution, which comprises: 3. A method for collecting radioactive metals by way of dissolution, which comprises: 4. A method according to claim 3, wherein the filtrate separated by filtration after electrolysis is treated with an ultraviolet irradiation device to further decompose organic matter contained in the filtrate. 5. A method according to claim 4, wherein the matter decomposed by irradiation of ultraviolet rays is separated by reverse osmosis from the filtrate. 6. A method according to claim 4, wherein the matter decomposed by irradiation of ultraviolet rays is separated by use of ion-exchange resin from the filtrate. 7. An apparatus for decomposing an organic solution containing radioactive metals, which comprises: 8. An apparatus for collecting radioactive metals, which comprises: 9. Apparatus for collecting radioactive metals according to claim 8, wherein a filtration device is disposed between the electrical conduction tank and the electrolytic device. 10. Apparatus for collecting radioactive metals according to claim 8 or 9, wherein the filtration device is provided with an ultraviolet irradiation device having a separation device, at a later stage thereof for ultraviolet irradiation and separation treatment of the filtrate therefrom. 11. Apparatus for collecting radioactive metals according to claim 10, wherein the separation device is of a reverse osmosis type. 12. Apparatus according to claim 10, wherein the separation device is of an ion-exchange resin type. |
description | 1. Field of the Invention The present invention relates to an X-ray radiographic apparatus and an X-ray radiographic method for radiographing a subject by detecting an X-ray applied from an X-ray tube to the subject, and more particularly to a technique for performing an X-ray radiography of the subject, employing a filter for extracting the X-ray components, and also relates to an X-ray restrictor used in the X-ray radiographic apparatus. 2. Description of the Related Art In related art, when the X-ray radiographic apparatus is employed to make the angiography for the lower limb, simultaneously for both legs, a contrast medium is injected into the blood vessel of the subject, and the contrast medium flowing within the blood vessel from the lumber region to the toe is radiographed with the X-ray radiographic apparatus. During this radiography, a holding member (e.g., C-type arm) for holding an X-ray tube and an X-ray detector (image intensifier) for the X-ray-radiographic apparatus, or a table top board for laying the subject thereon is being moved continuously or stepwise from the lumber region to the toe. However, the related-art X-ray radiographic apparatus had a problem that when the radiography is performed using a filter for extracting the X-ray components, the operator had a burden with the operation of the filter and the radiography becomes difficult. For example, when the angiography for the lower limb is made simultaneously for both legs, some halation occurs in the radiographed image because the transmission factor of X-rays is different between a leg portion and a portion between both legs. To prevent the halation, if a specific mat is placed between both legs of the subject, the operator has more burden with the mat. Also, the halation may be prevented by placing the filter near a bulb of the X-ray tube, in which the center of the radiographed image of the lumber region is hidden by the filter, making the radiography difficult. This invention has been achieved in the light of these affairs, and it is an object of the invention to provide an X-ray radiographic apparatus and X-ray radiographic method with high universality in which a burden of the operator is lightened, and also provide an X-ray restrictor used in the X-ray radiographic apparatus. In order to achieve the above object, this invention has the following constitution. An X-ray radiographic apparatus, according to a first aspect of the invention, comprising: an X-ray tube for radiating X-rays to a subject; an X-ray detector for detecting an X-ray applied to the subject; a holding member for holding the X-ray tube and the X-ray detector; a table top board for laying the subject thereon; a filter for extracting X-ray components radiated from the X-ray tube; and a movement mechanism for moving the filter outside or inside a radiation field of the X-rays radiated from the X-ray tube so that the filter is moved inside the radiation field, only when the relative position between the holding member and the table top board is at a predetermined position, and the filter is moved outside the radiation field, when the relative position is other than the predetermined position. With the invention of the first aspect, the movement mechanism moves the filter outside or inside the radiation field of the X-rays radiated from the X-ray tube, wherein the movement mechanism is configured such that the filter is moved inside the radiation field, only when the relative position between the holding member and the table top board is at a predetermined position, or the filter is moved outside the radiation field, when the relative position is other than the predetermined position. Accordingly, to acquire the radiographed image at the predetermined position employing the filter, the movement mechanism moves the filter inside the radiation field only at the predetermined position, thereby performing an X-ray radiography of the subject at the predetermined position through the filter. Further, to acquire the radiographed image other than at the predetermined position without employing the filter, the movement mechanism moves the filter outside the radiation field other than at the predetermined position, thereby performing an X-ray radiography of the subject other than at the predetermined position without the filter. Consequently, it is possible to lighten a burden of the operator who places a mat on the subject as performed in the related art, and further to select whether or not to use the filter depending on the radiographing position, thereby facilitating the radiography, whereby the X-ray radiographic apparatus has a higher universality. The filter is not specifically limited, as long as the X-rays having a specific energy, or the X-ray components, can be extracted from the X-ray flux that is different for each wavelength of the X-rays. For example, the filter may restrict the X-ray flux radiated from the X-ray tube in a predetermined shape. The “restrict in a predetermined shape” may involve setting a higher transmission factor of the X-rays within the predetermined shape and a lower or null transmission factor of the X-ray outside the predetermined shape, or setting a higher transmission factor of the X-rays outside the predetermined shape and a lower or null transmission factor of the X-ray within the predetermined shape. Further, the movement mechanism may include a switching member for selectively switching a plurality of filters into a certain filter to move the filter inside or outside the radiation field of the X-rays. In this manner, various filters may be employed to perform the X-ray radiography. In this specification, the inventions regarding an X-ray radiographic method, the X-ray tube for use in the X-ray radiographic apparatus, and the X-ray restrictor are also disclosed. (1) An X-ray radiographic method for performing an X-ray radiography of a subject, comprising: radiating X-rays from an X-ray tube to the subject; detecting an X-ray applied to the subject with an X-ray detector; moving at least one of a holding member for holding the X-ray tube and the X-ray detector and a table top board for laying the subject thereon; moving a filter, which extracts X-ray components radiated from the X-ray tube, outside or inside a radiation field of the X-rays radiated from the X-ray tube so that the filter is moved inside the radiation field, only when the relative position between the holding member and the table top board is at a predetermined position, or the filter is moved outside the radiation field, when the relative position is other than the predetermined position. With the invention (1), to acquire the radiographed image at the predetermined position employing the filter, the filter is moved inside the radiation field only at the predetermined position, thereby performing an X-ray radiography of the subject at the predetermined position through the filter, or to acquire the radiographed image other than at the predetermined position without employing the filter, the filter is moved outside the radiation field other than at the predetermined position, thereby performing an X-ray radiography of the subject other than at the predetermined position without the filter. Consequently, it is possible to lighten a burden of the operator who places a mat on the subject as performed in the related art, and further to select whether or not to use the filter depending on the radiographing position, thereby facilitating the radiography. (2) The X-ray radiographic method as defined in (1), may comprise selectively switching a plurality of filters into a certain filter to move the filter inside or outside the radiation field of the X-rays. With the invention (2), since the method includes selectively switching a plurality of filters into a certain filter to move the filter inside or outside the radiation field of the X-rays, various filters may be employed to perform the X-ray radiography. (3) An X-ray tube for use in an X-ray radiographic apparatus comprising an X-ray detector, a holding member for holding the X-ray tube and the X-ray detector and a table top board for laying a subject thereon, the X-ray tube comprising: an X-ray source for radiating X-rays to the subject; a filter for extracting X-ray components radiated from an X-ray source; and a movement mechanism for moving the filter outside or inside a radiation field of the X-rays radiated from the X-ray source so that the filter is moved inside the radiation field, only when the relative position between the holding member and the table top board is at a predetermined position, and the filter is moved outside the radiation field, when the relative position is other than the predetermined position. (4) An X-ray restrictor for controlling a radiation field of X-rays radiated from an X-ray tube of an X-ray radiographic apparatus, the X-ray radiographic apparatus comprising the X-ray tube, an X-ray detector, a holding member for holding the X-ray tube and the X-ray detector and a table top board for laying a subject thereon, the X-ray restrictor comprising: a filter for extracting X-ray components radiated from the X-ray tube; and a movement mechanism for moving the filter outside or inside the radiation field of the X-rays radiated from the X-ray tube so that the filter is moved inside the radiation field, only when the relative position between the holding member and the table top board is at a predetermined position, and the filter is moved outside the radiation field, when the relative position is other than the predetermined position. With the inventions (3) and (4), the movement mechanism is configured such that the filter is moved inside the radiation field, only when the relative position between the holding member and the table top board is at a predetermined position, or the filter is moved outside the radiation field, when the relative position is other than the predetermined position. Consequently, it is possible to lighten a burden of the operator, and realize the X-ray tube and the X-ray restrictor having a higher universality. (5) The X-ray tube as defined in (3), wherein the filter may restrict an X-ray flux radiated from the X-ray tube in a predetermined shape. (6) The X-ray restrictor as defined in (4), wherein the filter may restrict an X-ray flux radiated from the X-ray tube in a predetermined shape. The filter as defined in (3) and (4) is not specifically limited, like the filter of the first aspect of the invention, but may restrict the X-ray flux radiated from the X-ray tube in a predetermined shape, like the inventions (5) and (6). (7) The X-ray tube as defined in (3) or (5), wherein the movement mechanism may include a switching member for selectively switching a plurality of filters into a certain filter. (8) The X-ray restrictor as defined in (4) or (6), wherein the movement mechanism may include a switching member for selectively switching a plurality of filters into a certain filter. With the inventions (7) and (8), the movement mechanism is constituted of a switching member for selectively switching a plurality of filters into a certain filter to move the filter inside or outside the radiation field of the X-rays. In this manner, various filters may be employed to perform the X-ray radiography. One embodiment of the present invention will be described below with reference to the accompanying drawings. FIG. 1 is a side view showing the schematic constitution of an X-ray radiographic apparatus according to an embodiment of the invention, and FIG. 2 is a front view thereof. FIG. 3 is a view, partially broken away, of an X-ray tube and a collimator for use in the X-ray radiographic apparatus according to the embodiment of the invention, and FIG. 4 is a plan view thereof. FIGS. 5A and 5B are plan views of a switching mechanism equipped within the collimator. In this embodiment, an X-ray radiographic apparatus will be exemplified in which a blood vessel image is radiographed by injecting a contrast medium into the subject via a catheter for injecting the contrast medium. This X-ray radiographic apparatus comprises a table top board 1 for laying the subject thereon for medical examination, a strut portion 2, and a C-type arm 3, as shown in FIGS. 1 and 2. This strut portion 2 is attached to a ceiling plane and fixed except for rotation around the axial center of a vertical axis (z-axis in FIGS. 1 and 2). The C-type arm 3 corresponds to a holding member of the invention. The table top board 1 can be moved up and down and also moved along the body axis of the subject M. A table top board driving control section 4 is connected to the table top board 1 to control the driving of the table top board 1. The table top board driving control section 4 controls a driving mechanism for driving the table top board 1, a brake mechanism for stopping the movement of the table top board 1 in the body axis direction, a brake releasing mechanism for releasing the movement of the table top board 1 in the body axis direction, and a detection mechanism for detecting the position of the table top board 1 (not shown). The X-ray tube 5 is supported at one end of the C-type arm 3 to radiate the X-rays to the subject M. An image intensifier 6 (hereinafter appropriately abbreviated as “I.I”) is supported at the other end of the C-type arm 3 to detect the X-rays radiated on the subject M and convert them into an optical image. The C-type arm 3 can be rotated around the axial center of the body axis (x-axis in the FIGS. 1 and 2) of the subject M and also around the axial center of an axis (y-axis in FIGS. 1 and 2) in a perpendicular direction to the body axis of the subject M on the horizontal plane. The strut portion 2 can be rotated around the axial center of the vertical axis (z-axis in FIGS. 1 and 2). I.I 6 corresponds to an X-ray detector in this invention. A C-type arm driving control section 7 is connected to the strut portion 2 and the C-type arm 3, respectively to control the driving of the strut portion 2 and the C-type arm 3. The C-type arm driving control section 7 controls a drive mechanism for driving the strut portion 2 and the C-type arm 3 and a detection mechanism for detecting the position of the strut portion 2 and the C-type arm 3 (not shown). A high voltage generating section 8 is connected to the X-ray tube 5 to apply a bulb voltage and a bulb current to the X-ray tube 5. As shown in FIGS. 3 and 4, the X-ray tube 5 has a collimator 9 mounted on a plane in a direction for radiating the X-rays, in which this collimator 9 controls a radiation field of the X-rays. The collimator 9 comprises an X-ray movable restriction 10 having four plate-like members 10a and a switching mechanism 12 having a plurality of filters 11. The collimator 9 corresponds to an X-ray restrictor in this invention. Four plate-like members 10a constituting the X-ray movable restriction 10 are disposed orthogonal to each other to have an opening portion 10b, each plate-like member 10a being movable in the direction of the arrow a in FIGS. 3 and 4. The X-rays passing through the opening portion 10b become the radiation field of the X-rays, each plate-like member 10a being movable to adjust the size of the opening portion 10b and thereby the radiation field of the X-rays. As shown in FIGS. 1 and 2, a collimator drive control section 13 are connected to the plate-like members 10a and the switching mechanism 12, respectively to control the driving of the plate-like members 10a for the collimator 9 and the switching mechanism 12 (see FIGS. 3 and 4). The collimator drive control section 13 controls a drive mechanism (not shown) for driving the plate-like members 10a, a motor 22 for driving a base gear 16 of the switching mechanism 12, and a revolution counter 23 for detecting the position of the filter 11. A specific constitution of the switching mechanism 12 will be described later. The data converted into an optical image by I.I 6 is passed via a control section 14 to an arithmetical operation section 15 for performing various arithmetical operations, and outputting as a radiographed image. The control section 14 totally controls the table top board driving control section 4, the C-type arm driving control section 7, the high voltage generating section 8, and the collimator driving control section 13. The arithmetical operation section 15 has a function of calculating the relative position between the C-type arm 3 and the table top board 1, on the basis of the positions of the C-type arm 3 and the table top board 1, besides the arithmetical operations. Referring to FIGS. 3 to 5B, a specific constitution of the switching mechanism 12 will be described below. The switching mechanism 12 comprises the base gear 16 with a plurality of filters 11 disposed, a drive gear 17 and a position detecting gear 18 which are fitted with the base gear 16, a base shaft 19 passing through a central part of the base gear 16, a drive shaft 20 passing through a central part of the drive gear 17, a detection shaft 21 passing through a central part of the position detecting gear 18, a motor 22 for revolving the drive shaft 20 around the axial center, and the revolution counter 23 linked to the bottom of the detection shaft 21. In this embodiment, the filter 11 disposed in the base gear 16 has a lower limb radiographic filter 11a and three additional filters 11b, as shown in FIG. 4. The lower limb radiographic filter 11a restrict an X-ray flux radiated from the X-ray tube 5 to the quadrilateral shape with a quadrilaterally-shaped hole near the center thereof. The three additional filters 11b regulate the degree of transmission of X-rays. In this embodiment, the lower limb radiographic filter 11a is provided with a rectangular filter 11c with lower transmission factor of X-ray near the center, so that the X-ray flux transmitted through the lower limb radiographic filter 11a is restricted to the square shape with the rectangular hole near center thereof. In this embodiment, the filter 11c has the rectangular shape, but the filter 11c is not limited to the rectangular filter 11c. For example, the filter 11c may have an oval shape, which is disposed in the vicinity of the center in the lower limb radiographic filter 11a, corresponding to the portion between both legs. When the motor 22 revolves the drive shaft 20 around the axial center, the drive gear 17 is revolved, and the base gear 16 fitted with the drive gear 17 is also revolved. Each filter 11a, 11b disposed in the base gear 16 is revolved in the direction of the arrow b in FIG. 4 around the axial center of the base shaft 19 by the revolution of the base gear 16. Further, the revolution counter 23 measures the number of revolutions of the position detecting gear 18 fitted with the base gear 16 via the detection shaft 21, whereby the position of each filter 11a, 11b revolved in the direction of the arrow b in FIG. 4 is detected. Each filter 11a, 11b is moved outside or inside the opening portion 10b of the X-ray movable restriction 10, namely, outside or inside the radiation field of the X-rays by revolving the base gear 16, as shown in FIG. 4. A certain filter 11 is selected from among the four filters (lower limb radiographic filter 11a, additional filters 11b) by revolving the base gear 16 every 90° and switched. Accordingly, the switching mechanism 12 corresponds to a movement mechanism in this invention, as well as switching member in this invention. Referring now to a flowchart of FIG. 6, an X-ray radiographic method involving the angiography for lower limb will be described below. In this embodiment, an x-ray radiography is performed in the order from the lumber region to the toe with the contrast medium flowing inside the blood vessel. (Step S1) Movement of the Table Top Board After the subject M is laid on the table top board 1, the table top board is moved up or down with the subject M laid thereon, so that the table top board 1 is positioned at a predetermined height. Then, the table top board 1 is moved along the direction of the body axis of the subject M, so that the lumber region of the subject M is located-between the X-ray tube 5 and I.I 6 for the C-type arm 3. (Step S2) Radiography of the Lumber Region If the lumber region of the subject M is located between the X-ray tube 5 and I.I 6 for the C-type arm 3, it is started to radiograph the lumber region. Then, the additional filters 11b other then the lower limb radiographic filter 11a are located within the radiation field of X-rays, namely, the lower limb radiographic filter 11a is located outside the radiation field, whereby the lumber region is radiographed by transmitting the X-rays through the additional filters 11b inside the radiation field, as shown in FIG. 5A. More specifically, the X-rays are radiated from the X-ray tube 5, and via the additional filters 11b and the X-ray movable restriction 10 to the subject M. The X-rays applied to the subject M are detected and converted into an optical image by I.I 6. The converted data is transferred via the control section 14 to the arithmetical operation section 15 for performing various arithmetical operations to acquire a radiographed image of the lumber region. In this embodiment, since it is intended to examine how the contrast medium flows inside the blood vessel, an output result of the radiographed image is displayed on a TV monitor (not shown) and examined without actually acquiring the radiographed image. The table top board 1 is moved in a direction of the head of the subject M, to acquire the radiographed image from the lumber region to the toe, while examining the radiographed image of the lumber region. The radiographed image is acquired in the order while the table top board 1 is being moved. (Step S3) Switch of the Filter If the groin (crotch of the thigh) of the subject M reaches the position between the X-ray tube 5 and I.I 6, the switching mechanism 12 switches the filter 11. That is, the relative position between the holding member and the table top board in this invention means the relative position between the C-type arm 3 and the table top board 1 where a region of the subject M on the toe side from the groin is located between the X-ray tube 5 and I.I 6 for the C-type arm 3. In switching, the additional filters 11b inside the radiation field are revolved in a direction of the arrow in FIG. 5B, and moved outside the radiation field as shown in FIG. 5B. Instead, the lower limb radiographic filter 11a, which is located outside the radiation filed, is revolved in the same direction and moved inside the radiation field as shown in FIG. 5B. Thereby, the lower limb radiographic filter 11a is selected and switched. The revolution counter 23 detects the position of the filter 11, whereby the lower limb radiographic filter 11a is controlled to be located inside the radiation field. (Step S4) Radiography of Both Legs After the lower limb radiographic filter 11a is switched, it is started to radiograph both legs. In radiographing both legs, the X-rays are passed through the lower limb radiographic filter 11a inside the radiation field. More specifically, of an X-ray flux radiated from the X-ray tube 5, the X-rays near the center are hardly transmitted by the rectangular filter 11c near the center of the lower limb radiographic filter 11a, and the X-rays in other portions are transmitted. Thereby, the X-ray flux transmitted through the lower limb radiographic filter 11a is applied to the subject M in a restricted state of square shape with the rectangular hole near the center thereof. Accordingly, the X-rays are applied to a portion of the leg of the subject M corresponding to the position of square shape of the lower limb radiographic filter 11a, but hardly applied to a portion between both legs corresponding to the vicinity of the center of the lower limb radiographic filter 11a, or the rectangular filter 11c. The radiographed image of both legs is displayed on the TV monitor and examined, on the basis of the data of the X-rays applied to both legs of the subject M in the same manner as radiography of the lumber region in step S2. The table top board 1 is moved in a direction of the head of the subject M, to acquire the radiographed image from the groin to the toe, while examining the radiographed image of both legs. The radiographed image is acquired in the order while the table top board 1 is being moved. If the radiographed image is acquired up to the toe, the table top board 1 with the subject M laid thereon is moved along the body axis of the subject M to avoid the C-type arm 3. After the table top board 1 is moved up or down, the subject M is let off the table top board 1, whereby the series of X-ray radiography is ended. In the series of X-ray radiography according to the steps S1 to S4, to acquire the radiographed image of both legs using the lower limb radiographic filter 11a in step S4, the lower limb radiographic filter 11a is moved inside the radiation field in step S3, only when both legs are radiographed, thereby performing an X-ray-radiography of both legs of the subject M through the filter 11a. To acquire the radiographed image of the lumber region in step S2, other than both legs, without employing the filter 11a, the filter 11a is moved outside the radiation field, thereby performing an X-ray radiography of the lumber region of the subject M without the filter 11a. In the apparatus of this embodiment, the switching mechanism 12 is configured to move the filter 11 outside or inside the radiation field of X-rays radiated from the X-ray tube 5. Also, when the relative position between the C-type arm 3 and the table top board 1 is at the predetermined position, where a region of the subject M on the toe side from the groin is located between the X-ray tube 5 and I.I 6 for the C-type arm 3 in this embodiment, the lower limb radiographic filter 11a of the four filters is moved inside the radiation field, while the relative position is not at the predetermined position, the filter 11a is moved outside the radiation field. In this embodiment, when radiographing both legs, the X-rays are applied to a portion of the leg of the subject M corresponding to the position of square shape of the lower limb radiographic filter 11a, but hardly applied to a portion between both legs corresponding to the vicinity of the center of the lower limb radiographic filter 11a, or the rectangular filter 11c, reducing a difference in the transmission factor between the leg portion and the portion between both legs. Consequently, no halation occurs, whereby there is no need for placing a mat over the subject M to prevent the halation, and it is possible to lighten a burden on the operator. Further, when radiographing the lumber region of the subject M, for example, other than both legs, the additional filters 11b are employed in place of the lower limb radiographic filter 11a, so that the center of the radiographed image of the lumber region is not hidden by the rectangular filter 11c in the lower limb radiographic filter 11a. Consequently, it is possible to select whether or not to employ the lower limb radiographic filter 11a depending on the radiographing position, facilitating the radiography, whereby the X-ray radiographic apparatus with high universality is realized. In this embodiment, the switching mechanism 12 has means for selecting and switching the lower limb radiographic filter 11a from among the four filters 11, whereby the X-ray radiography is made using various filters. This invention is not limited to the above embodiment, but may be modified or varied in the following manner. (1) In the above embodiment, a contrast medium is injected via a contrast medium injecting catheter, and it is examined how the contrast medium flows inside the blood vessel. However, this invention is also applicable to a digital subtraction angiography in which the radiographed images before and after catheter injection are subject to a subtraction operation. Also, this invention is applicable to an X-ray radiographic apparatus for simply performing the X-ray radiography for medical examination without catheter injection, and an X-ray CT apparatus. (2) In the above embodiment, the filters 11 to be switched are the lower limb radiographic filter 11a and the additional filters 11b, but not limited to those filters 11a and 11b. Also, the lower limb radiographic filter 11a is only provided in the base gear 16, but when the filter 11a is not employed, the X-ray radiography may be made without any use of other filters. (3) In the above embodiment, the switching mechanism 12 is employed as movement mechanism of this invention, and has the function of the switching member in this invention, but the switching member is not limited to the switching mechanism 12 in this embodiment. For example, a plurality of filters 31 are disposed on the belt 30, as shown in a cross-sectional view of FIG. 7, and the filters 31 are selected and switched by moving the belt 30 in a direction of the arrow in FIG. 7. Also, it does not necessarily follow that the function of switching member is provided. The lower limb radiographic filter 11a may be moved outside or inside the radiation field by providing the lower limb radiographic filter 11a alone in the base gear 16 as described in the variation example (2). Alternatively, a single filter 41 may be disposed on a base board 40 movable in the horizontal direction, in place of the base gear 16 of this embodiment, whereby the filter 41 may be moved outside or inside the radiation field by moving the base board 40 in the horizontal direction, as shown in a plan view of FIG. 8. (4) In the above embodiment, the switching mechanism 12 as the movement mechanism in this invention is provided in the collimator 9 that is the X-ray restrictor in this invention, but the movement mechanism in this invention may be provided in the X-ray tube 5. (5) In the above embodiment, the holding member in this invention uses the C-type holding member 3 with the C-type arm 8, but is not specifically limited as long as the holding member holds the X-ray tube and the X-ray detector. As will be apparent from the above description, with the invention of the first aspect, the movement mechanism is configured such that the filter is moved inside the radiation field, only when the relative position between the holding member and the table top board is at a predetermined position, or the filter is moved outside the radiation field, when the relative position is other than the predetermined position. Consequently, it is possible to lighten a burden of the operator, whereby the X-ray radiographic apparatus has a higher universality. |
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abstract | An x-ray collimator can be constructed from multiple subassemblies, which at least includes a first subassembly that reduces the leakage of x-ray radiation between adjacent apertures and a second subassembly that reduces the spill of x-ray radiation around the detector face. Each of these subassemblies has numerous apertures. In the first subassembly these apertures correspond to focal spots on an x-ray source, and in the second subassembly, these apertures are shaped such that the dimensions increase from smaller entrances to larger exits. |
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abstract | A method for generating x-ray radiation, comprising the steps of forming a target jet by urging a liquid substance under pressure through an outlet opening, the target jet propagating through an area of interaction; and directing at least one electron beam onto the target jet in the area of interaction such that the electron beam interacts with the target jet to generate x-ray radiation; wherein the full width at half maximum of the electron beam in the transverse direction of the target jet is about 50% or less of the target jet transverse dimension. A system for carrying out the method is also disclosed. |
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claims | 1. A method of manufacturing a dispersion-ceramic micro-encapsulated (DCM) nuclear fuel comprising:forming two or more different sizes of tristructural-isotropic (TRISO) fuel particles, wherein the two or more different sizes of TRISO fuel particles each include a fuel kernel surrounded by a porous carbon buffer layer, an inner pyrolytic carbon layer, a ceramic layer, and an outer pyrolytic carbon layer; andobtaining a fuel compact of DCM fuel that includes the two or more different sizes of tristructural-isotropic (TRISO) fuel particles dispersed and embedded inside a silicon carbide matrix by:mixing the two or more different sizes of TRISO fuel particles with a silicon carbide matrix precursor material, the silicon carbide matrix precursor material including a silicon carbide powder;sintering the silicon carbide matrix precursor material and the two or more different sizes of TRISO fuel particles through a nano infiltration transient eutectic phase to provide filling of the space between the two or more different sizes of TRISO fuel particles by pressing in a mold to stress the silicon carbide based precursor material and the two or more different sizes of TRISO fuel particles mixture at a predetermined sintering pressure and temperature to disperse and embed the two or more different sizes of TRISO fuel particles inside the silicon carbide matrix, thereby obtaining the fuel compact of DCM nuclear fuel. 2. The method of claim 1, further comprising:overcoating an entirety of the TRISO fuel particles with an additional silicon carbide layer prior to sintering. 3. The method of claim 2, wherein the mold is a cylindrical shape to form the fuel compact of DCM nuclear fuel as a cylindrical pellet, and the method further comprises:decreasing a pellet-to-clad gap of the fuel compact and an intended cladding tube for stacking the cylindrical pellet inside by increasing a diameter size of the cylindrical pellet. 4. The method of claim 2, further comprising:providing a burnable poison in a fuel pin of a plurality of fuel compacts of the DCM nuclear fuel by a presence of resonant absorbers. 5. The method of claim 4, wherein the resonant absorbers are selected from a group consisting of gadolinium or erbium. 6. The method of claim 1, wherein the two or more different sizes of TRISO fuel particles are formed by:dissolving material for the fuel kernel in a nitric acid to form uranyl nitrate solution;dropping the uranyl nitrate solution through a nozzle to form droplets or microspheres;gelling and calcining the droplets or microspheres to produce the fuel kernel; andrunning the fuel kernel through a coating chamber to sequentially coat the fuel kernel with the porous carbon buffer layer, the inner pyrolytic carbon layer, the ceramic layer, and the outer pyrolytic carbon layer. 7. The method of claim 6, wherein the coating chamber is a chemical vapor deposition furnace. 8. The method of claim 6,wherein the silicon carbide powder has an average size of less than 1 μm and a specific surface area greater than 20 m2/g. 9. The method of claim 8, wherein the silicon carbide powder ranges from 15 nm to 51 nm. 10. The method of claim 8, wherein the silicon carbide powder has a mean particle size of 35 nm. 11. The method of claim 8, further comprising:during or prior to mixing the two or more different sizes of TRISO fuel particles with the silicon carbide matrix precursor material, adding sintering additives to the silicon carbide powder or coating the sintering additives onto the silicon carbide powder. 12. The method of claim 11, wherein the sintering additives range from 6 weight % to 10 weight % of the silicon carbide powder. 13. The method of claim 8, wherein mixing the two or more different sizes of TRISO fuel particles with the silicon carbide precursor material comprises:mixing the two or more different sizes of TRISO fuel particles with the silicon carbide powder precursor material that includes the silicon carbide powder in a liquid slurry state or a powder state. 14. The method of claim 1, wherein the predetermined sintering pressure and temperature during the pressing is less than 30 MPa and 1900° C. 15. The method of claim 1, wherein the predetermined sintering pressure and temperature during the pressing is 10 MPa and 1850° C. 16. The method of claim 1, wherein a duration of the pressing is less than or equal to one hour. 17. The method of claim 1, wherein a duration of the pressing is more than one hour. 18. The method of claim 1, wherein the silicon carbide matrix has a closed microporosity of 3% to 4%. 19. The method of claim 1, wherein the two or more different sizes of TRISO fuel particles includes a first TRISO fuel particle having a first fuel kernel with a first kernel radius of no more than 375 micrometers and a first packing fraction of no more than 45% and a second TRISO fuel particle having a second fuel kernel with a second kernel radius of no more than 200 micrometers and a second packing fraction of no more than 3%. |
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description | This Application claims priority from U.S. Provisional Application 61/390,729, filed Oct. 7, 2010, which is hereby incorporated by reference. The invention relates to a beam blanker for interrupting a beam of charged particles, the beam blanker showing an axis along which charged particles propagate before entering the beam blanker, the beam blanker comprising means for generating an electric field perpendicular to said axis, the electric field for deflecting the charged particles, and an aperture in a diaphragm, the aperture transmitting the beam when the beam is not interrupted and the diaphragm stopping the beam when the beam is interrupted. Such a beam blanker is known from U.S. Pat. No. 4,445,041. Beam blanking is used in for example charged particle instruments, such as Scanning Electron Microscopes, Transmission Electron Microscopes, Focused Ion Beam machines, and the like. A well-known application is the use of a beam blanker for modulating the beam current in an Electron Beam Lithography System. In such a system a beam of energetic electrons is focused and scanned over a substrate, such as a semiconductor wafer covered with a resist layer. During this scanning the beam is blanked/unblanked, as a result of which part of the resist layer is irradiated by the electron beam and part of the resist layer is not exposed to electrons. Due to the irradiation the properties of the resist are locally changed, and further treatment of the resist results in the forming of a pattern on the wafer. Further processing, such as etching, evaporation of metals, implantation of materials, etc, may then be performed. The known beam blanker is intended to blank/unblank a beam of electrons as used in such an Electron Beam Lithography System. It comprises a deflector formed by two electrodes and downstream of said deflector a diaphragm with an aperture. The diaphragm intercepts the beam of electrons when the beam is deflected by a deflector, and the aperture transmits the beam when the beam is not or hardly deflected. For a fast blank/unblank speed the diaphragm is positioned at a cross-over position of the beam. The small spot size of the beam at the cross-over enables an abrupt change in current for a small change in deflection, and thus a fast blank/unblank speed for a given dV/dt of the deflection signal. Further downstream of the deflector and the diaphragm a second deflector is placed so that the combined deflection of the two deflectors results in a pivot point at the cross-over. As a result of the pivot point coinciding with the cross-over no change of the position of the spot is observed downstream of the beam blanker when blanking the beam. Due to the limited speed of electrons in a beam the deflectors should be excited with a small time delay. This is achieved by exciting the deflectors with the same signal, but the signal of the second deflector is delayed by adding a delay line. The voltage needed to blank the beam is typically 6 volts at a beam energy of approximately 20 keV. According to the patent disclosure the known beam blanker can be used to frequencies of around 300 MHz. Another application of a beam blanker is for generating a pulsed beam in a Transmission Electron Microscope (TEM). In a TEM a sample is irradiated by a beam of energetic electrons, and electrons transmitted through the sample are used to gather information about the sample. Normally the beam is used to study a sample that does not change in time, or only very little. Lately there is a demand for studying effects that are time dependent, such as decay effects after probing the sample with a pulse of light, thereby pumping it to an excited state. By irradiating the sample with a train of electron pulses while also illuminating the sample with a time-synchronized train of light pulses from a pico- or femto-second laser, resulting in a train of light pulses with a duration similar to or shorter than the electron pulse, the decay effects can be studied. By introducing a variable phase delay between the two trains of pulses and making recordings for different phase delays, a complete dependency of the decay effects can be recorded. This resulted in the 1999 Nobel Prize for chemistry to professor Zewail for his work on femtochemistry, that is: the study of chemistry in the femtosecond scale by observing molecules with ultra-fast lasers. It is noted that Zewail did not use a beam blanker to generate a pulsed electron beam, but a pulsed electron source, as described in US patent application No. US2005/0253069. The application describes that an electron source (a heated LaB6 crystal) is pulsed by photo-emission of using a pulsed laser. A problem of a pulsed electron source of the type used by Zewail is that it shows a brightness that is much lower than the brightness of the well-known Schottky emitters routinely used in TEMs. As known to the person skilled in the art the amplitude of the signal for driving a beam blanker is proportional to the energy of the beam blanked, while the period of the signal (and thus the dV/dt) is proportional with the frequency and/or rise time of the beam. At increasing frequencies the signal for driving the beam blanker must thus have a larger dV/dt. This is difficult to achieve. This is aggravated when the blanker is used for blanking a beam with higher energy, as this implies (using the same dimensions) a blanking signal with increased amplitude and increased dV/dt. It is noted that the 300 keV beam energy typically used in a TEM is more than 10 times higher than the 20 keV used in the beam blanker of the known patent. There is a need for a high frequency beam blanker operating at a lower power and/or with a higher sensitivity. The invention intends to provide such a beam blanker. To that end the beam blanker according to the invention is characterized in that the electric field is generated by a resonant structure with a resonant frequency f, the resonant structure equipped to generate an electric field that sweeps the beam over the aperture, as a result of which the beam is transmitted through the aperture twice per period of the frequency f. By making the deflection means part of a resonant structure, the amplitude of the deflector is amplified by a factor Q, in which Q is the quality factor of the resonant structure. Such a resonant structure may include an LC network. In a preferred embodiment the deflector means consist of two electrodes that are part of a capacitor of an LC network. Alternatively the deflector means are part of a wave guide, the wave guide coupled to a RF oscillator. The waveguide may be open or closed at one end. Small apertures in the waveguide allow the beam to enter and to leave the wave guide. The electric field in the waveguide deflects the beam. In a preferred embodiment the resonant structure comprises a resonant transmission line (shielded or non-shielded). As is the case for a wave guide, the transmission line may be open ended or closed, as long as the beam passes near a voltage maximum of the transmission line. It is noted that, as known to the person skilled in the art of RF electronics, when driving the resonant structure by a generator, impedance matching may be done by, for example, stub tuning with one or more opened or closed stubs. At lower frequencies for example tapped inductors and/or capacitors may be used, or for example a helical resonator that may or may not be combined with a capacitor. It is noted that a beam blanker is described by K. Ura et al., “Picosecond Pulse Stroboscopic Scanning Electron Microscope”, J. Electron Microsc., Vol. 27, No. 4, (1978), p. 247-252. This beam blanker uses a resonant cavity as shown in its FIG. 2a, and a buncher as shown in its FIG. 2b. The document mentions that the cavities are designed using Fujisawa's theory, see K. Fujisawa, “General Treatment of Klystron Resonant Cavities”, IRE Trans. on Microwave Theory and Techniques (October 1958), pages 344-358. It is noted that the theory of Fujisawa only relates to (klystron) cavities with rotational symmetry, and thus the cavities of Hosokawa are rotational symmetric as well. Such cavities do not generate an electric field perpendicular to the rotational axis, as is the case in this invention. In a preferred embodiment the beam blanker comprises a resonant transmission line and a grounded conductor, and the electric field is generated between the resonant transmission line and the grounded conductor. The resonant transmission line may be open or closed at the end, and a voltage maximum (a voltage node) occurs at the position where the beam passes the transmission line. The electric field may be synchronized to, or derived from, a driving signal. The driving signal may be an electric signal, or it may be an optical signal for triggering a photoreceptor, such as a PIN diode or a phototransistor, in an electric circuit. Synchronizing the signal with an optical signal is especially attractive when synchronizing the signal to an optical probe signal that probes the sample in, for example, a TEM. Synchronizing the signal with an optical signal is also very attractive when the beam blanker is situated in a high voltage area, such as the gun area of a TEM. The deflection of the beam depends on the energy of the charged particles. In charged particle instruments the particles are often generated in a gun area and accelerated to their final energy The energy in the gun area is typically lower than 10 keV, while in a Transmission Electron Microscope the final energy with which they impact on a sample is typically between 80 and 400 keV, although higher and lower energies are known to be used. By placing the beam blanker in the gun area the deflection voltage (and thus the power needed to drive the beam blanker) is less than when the beam blanker is operated at the final energy. Preferably the driving of such a signal at high voltage (the gun voltage) is done using previously mentioned triggering by an optical signal, in which the optical signal bridges the gap between ground to high voltage via a fiber. The resonant frequency f may be derived from a driving signal by injection locking, phase locking or frequency locking the resonant frequency to a, in most cases sub-harmonic, driving signal. Also frequency multiplication may be used to form a high frequency signal from a lower frequency signal. Instead of driving the resonant structure with a driving signal, the resonant structure may comprise a negative impedance element, such as a Gunn diode or an IMPATT diode. In this way no external driving signal is necessary, as the negative impedance element will make the structure oscillate. It is noted that circuits are known in which the frequency of such a circuit can be tuned in a variety of ways, including mechanical means, electronic means (phase shifters) or injection locking. For some uses it is necessary to tune frequency and/or phase of the resonant frequency f. This may be achieved by, for example, mechanical tuning means (e.g. a tuning screw), by electronic means (e.g. a phase shifter or a varicap), or in another way (such as the magnetic tuning used in RF isolators/circulators. It is noted that the latter (the RF isolator/circulator) may also ease the demands on impedance matching to the resonant structure as little energy is reflected back into the circuitry generating the driving signal. Instead of using a transmission line in the resonant circuit, also a cavity resonator may be used to generate the electric field. The cavity resonator may take the form of a TM or TE waveguide with two holes through which the beam enters and leaves the cavity resonator or waveguide. Preferably the beam blanker is equipped with an aperture in the form of a slit or a hole with a dimension in the direction in which the beam is deflected of less than 100 μm, preferably less than 10 μm, most preferably less than 1 μm. In a particle-optical instrument such as a Scanning Electron Microscope (SEM) or a Transmission Electron Microscope (TEM), or an instrument equipped with a SEM and/or a TEM column, the electron source is often a Schottky emitter. The Schottky emitter typically has an emitting diameter of approximately 20 nm, and a cross-over of less than 20 nm is formed by the condenser optics. An aperture of 100 μm, preferably less than 10 μm, and most preferably less than 1 μm can thus be well used to transmit the beam. It is noted that a smaller aperture implies that the beam is blanked at a lower electric field (a smaller deflection), but that too small an aperture implies problematic alignment. However, an alignment of less than 1 μm is well achievable. The beam blanker is for use in a particle-optical instrument such as a Scanning Electron Microscope (SEM), a Transmission Electron Microscope (TEM) or a Focused Ion Beam (FIB) instrument, or an instrument equipped with a SEM, FIB and/or a TEM column. Preferably the apparatus is equipped with a laser, such as a nano- or femtosecond laser, producing a train of light pulses for probing the sample. When synchronizing the laser and the beam blanker, time dependent studies on ultra-short (femto-second) scale or longer can be performed. The synchronization can be achieved by e.g. triggering the beam blanker by a laser pulse. Electrons then irradiate the sample shortly after the triggering took place. A variable phase shift in the beam blankers circuitry can then be used to probe the sample at different delay times. For longer delay times two beam blankers can be used, one of the beam blankers selecting some of the pulses transmitted by the other. Preferably the two blankers show positional overlap, or even share components, such as the aperture. As is clear to the person skilled in the art the signal of the beam blankers should be such that the beam passes through both blankers when required. This can be achieved by driving the beam blankers with the same frequency, but varying the phase of one (or both) signals with a phase shifting element. It is also possible to excite one beam blanker with a first frequency and the other with a frequency that has a harmonic or sub-harmonic relation to the first frequency. In that case a zero-crossing of both frequencies, and thus a passing of the beam through both beam blankers, occurs on a regular basis. FIG. 1 schematically shows a TEM equipped with a beam blanker according to the invention. A source of charged particles in the form of electron source 102 emits a beam of electrons 104a round electron-optical axis 106. An electron-optical lens 108, in the form of a magnetic or an electrostatic lens, focuses the beam in the beam blanker 110. A signal generator 112 is connected to the beam blanker to provide a driving signal to the beam blanker. The beam 104b leaving the beam blanker enters two condenser lenses 114 and 116, and the opening angle of the beam is limited by beam limiting aperture 118. Thereafter the beam illuminates the electron transparent sample 120. The sample is mounted on a sample holder 122 that may shift or tilt the sample. The so-called objective lens 124 forms a first magnified image of the sample, that is further magnified by lenses 126 and 128 to form an image on image plane 130. The image plane may coincide with a fluorescent screen, a CCD camera or a CMOS camera, or it may coincide with the entrance plane of another type of detector. It is noted that TEMs with more condenser lenses, and/or more imaging lenses are known. Furthermore a TEM also comprises dipoles for aligning the beam, and may comprise multipoles for correcting e.g. lens errors. Other detectors, for example detectors for detecting radiation from the sample in the form of back-scattered electrons, secondary electrons, X-rays, etc. may be used. The energy loss of electrons transmitted through the sample may be determined. The source of particles typically used in a TEM is a Schottky source, although other sources, such as thermionic sources comprising a crystal of e.g. LaB6 or CeB6 are also known to be used, as well as field-emission sources. The electrons emitted by such a source are accelerated to an energy of typically between 50 keV and 500 keV before impinging on the sample, although lower and higher energies are known to be used. The sample is a very thin sample, ranging from from a thickness of less than 30 nm when the sample comprises high-Z materials, such as semiconductor samples, to samples of 1 μm when the sample comprises mainly low-Z materials, as is the case for biological and e.g. polymer samples. Because the sample is so thin, and the electrons are so highly energetic, the sample is transparent to the electrons. However, there is interaction between the electrons and the sample, as a result of which electrons may be scattered, or absorbed. Scattering of the electrons may be detected by constructive/destructive interference of the scattered electrons with non-scattered electrons on image plane 130. Absorption of electrons may likewise be imaged on the enlarged image plane. In this manner images can nowadays be made with atomic resolution (less than 0.1 nm) and a magnification of more than 1,000,000 times. Normally images are obtained of a sample that does not, or hardly, change during the imaging. In most cases it is even detrimental when sample change (either in appearance or in position) during the imaging. However, some studies comprise the time dependent imaging of samples that are probed by e.g. a laser pulse. Such studies may address for example solidification studies of a sample, or the decay of excited states. Essential herein is synchronization between the bunching of the beam blanker and the probing of the sample with e.g. the laser pulse. By using the beam blanker many bunches of electrons are leaving the beam blanker, and thus the sample is illuminated by regularly repeated bunches of electrons. Even when one bunch of electrons does not contain sufficient electrons to form an image with sufficient quality (for example signal-to-noise ratio), repetition of bunches with a constant time delay between laser pulse and electron bunch may result in an image with sufficient quality. By slightly changing the time delay a series of images corresponding to different time delays can be formed. It is noted that the use of a Schottky-emitted electron source or the use of a field-emitter electron source results in an electron source with a much higher brightness than the electron source used by Zewail (thermionic LaB6 crystal with pulsed laser photo-emission). It is further mentioned that the electrons can be accelerated to their final energy (the energy with which they impinge on the sample) before traveling through the beam blanker or after traveling through the beam blanker. FIG. 2 schematically shows a Scanning Electron Microscope equipped with a beam blanker according to the invention. FIG. 2 shows an electron source 102 emitting an electron beam 104a round an electron-optical axis 106. An electron-optical lens 108, in the form of a magnetic or an electrostatic lens, focuses the beam in the beam blanker 110. A signal generator 112 is connected to the beam blanker to provide a driving signal to the beam blanker. Objective lens 202 focuses beam 104b coming out of the beam blanker on the sample 204. The beam is scanned over the sample by deflectors 206a and 206b. The sample is mounted on a stage 208 that may shift or tilt the sample. The electrons impinging on the sample cause secondary radiation 210, such as backscattered electrons, secondary electrons, photons, X-rays, etc. Such secondary radiation may be detected by detector 212. It is worth mentioning that also detectors are known that detect radiation that re-enters the objective lens. In that case the detector is typically situated between the objective lens and the condenser lens 116. It is noted that similar instruments are known in which, instead of electrons, ions are used as charged particles. It is further noted that scanning of a focused beam over the sample is also known in Scanning Transmission Electron Microscopes (STEMs). Such STEMs are often equipped to work as a TEM as well. FIG. 3 schematically shows a beam blanker 100 according to the invention. The beam blanker comprises two deflector electrodes 308 and 310 for generating an electric field. The incoming beam of particles 104a, propagating along axis 300, passes through the aperture 304 in diaphragm 302 when not deflected. When deflected, the beam is intercepted by the diaphragm. Electrically the electrodes 308 and 310 form a capacitor that is in parallel to inductor 312, thus forming a resonant circuit. The inductor is a tapped inductor, and thus the amplitude of signal generator 314 is greatly magnified. As a result bunches of charged particles 306-i leave the beam blanker. In this figure the resonant circuit is shown as a LC circuit, with lumped components. For high blanking frequencies, for example frequencies in excess of 1 GHz, more specifically in excess of 10 GHz, resulting in bunch length of 1 ps or less, the use of transmission lines is more appropriate. As known to the person skilled in the art of RF electronics, a transmission line that is not terminated with its characteristic impedance can be a resonant transmission line where at one or more positions on the transmission line the voltage of a signal fed to the transmission line shows a maximum, for example at λ/4 from the end of a shorted transmission line, with λ the wave length of the electric signal. For transmission lines that are open at the end, the maximum occurs at the end and, for example, at λ/2 from the end. By passing the beam near such a maximum, high deflection is achieved with low driving power of the transmission line. Tuning may be done with one or more stub tuners. It is noted that a structure that is tuned at a frequency f, often is also tuned at either odd or even harmonics of said frequency. It is mentioned that, assuming a voltage maximum occurs at given location on the transmission line, also positions removed much less than λ/4 from said position can be used to generate the electric field. It is thus sufficient when the beam passes, for example, within 1/10, more preferably 1/20, of a wavelength of a voltage maximum. A position close to an open-ended transmission line is thus suited. At a resonant frequency of approximately 25 GHz a wavelength corresponds with approximately 1 cm, and thus positional accuracies of approximately 1 mm are needed. It is noted that preferably the length of the beam blanker, that is: the dimension in the direction of the propagation direction of the beam, is small, so that the lingering period of the charged particles in the beam blanker is small compared to the bunch length. A preferred embodiment of the beam blanker has a length in the direction in which the beam propagates such that a particle of the beam resides in the electric field less than 1/10 of a period of the resonant frequency, more preferably less than 1/100 of a period of the resonant frequency. By using two beam blankers, one after the other, a number of bunches 306-i can be removed from the train of bunches: a charged particles passes through the two beam blankers when it experiences no deflection at both blankers, and is blocked when it experiences a deflection at either blanker. Reduction of the number of pulses can be desired for studying longer lasting phenomena, for phenomena in which it is desired that the sample returns to its ground state, or in cases where heating by the electron beam is a problem. Control can be realized by feeding a signal with the same frequency to both blankers, but shifting the phase of the signal fed to one of the blankers, for example using an electronically controlled phase shifter as commercially available, or a phase shifting element such as a varicap. Another way of control is by feeding a signal having a harmonic or sub-harmonic relation of the signals fed to the two blankers, resulting in an eye-pattern that crosses the origin. It is noted that, as the traveling speed of the particles is limited, a proper delay time between the two beamblankers should be included. It is noted that skipping a number of pulses may look identical to lowering the resonant frequency f to a sub-harmonic. However, using a lower resonant frequency will normally also result in a longer pulse length, as—assuming the same amplitude for the electric field—the beam takes longer to pass over the aperture. The use of a lower resonant frequency may thus result in a degradation of time resolution. FIG. 4 schematically shows the beam position of the beam on the diaphragm 302 when using two blankers and signals with harmonic relation. Assuming the beam blankers deflect the beam 104a in two perpendicular directions, the beam will be deflected or scanned over the diaphragm in a Lissajous-figure 402 as shown in FIG. 4. Only when the beam is over the aperture 304 the beam passes through the aperture, otherwise it is blocked by diaphragm 302. It is noted that a proper delay time between the two beam blankers should be included to correct for the traveling time of the particles between the two blankers. It is further noted that the beam blankers may share components, such as the diaphragm 302. In a Transmission Electron Microscope (TEM) the beam of charged particles is produced at an energy of, for example, 5 keV and then accelerated to an energy of typically between 50 keV and 500 keV. For an identical behavior of the blanker the voltage on the blanker should be proportional to the energy of the charged particles (the electrons). A much lower voltage can thus be used when blanking the beam before it is accelerated to its final energy. In such a case the synchronization of the beam blanker to for example a laser probing the sample is preferably done by sending (a derivative of the) light pulses synchronized with the laser via an optical path to photo-receptors at high-voltage. These photo-receptors can then be part of the electronics that drive the blanker or blankers. It is noted that, as a result of the lower power needed for a resonant structure, electrical interference, both internal to the instrument and external to the instrument, are smaller. This makes this type of beam blanker well suited for generating pulses with a repetition frequency of, for example, between 10 GHz and 100 GHz, and pulses with a duration of 10 ps or less. When used in a transmission electron microscope equipped with a high-brightness electron source, for example a Schottky emitter, such a blanker is suited for the study of chemistry on the picoseconds timescale. |
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041464230 | summary | The invention relates to a nuclear reactor having a water-cooled core, a pressure vessel of pre-stressed concrete and an inner containment vessel spaced from the pressure vessel so that an insulating gap is formed between the pressure vessel and the containment. The invention relates specifically to a water-cooled reactor of the type described, in which the pressure vessel contains both hot cooling water and steam, preferably a boiling water reactor. In a nuclear reactor of the type described it is desirable for the temperature of the concrete not to exceed about 50.degree. C. This is achieved primarily by said insulating gap which may be filled with water or with a gas. Insulation is improved if an insulating layer is applied on the inner containment. A cooling system is also arranged in the concrete pressure vessel, suitably in the form of cooling pipes in the concrete close to the inner wall of the pressure vessel. The object of the invention is to effect a reactor in which the above-mentioned inner containment is divided into two parts which are not secured to each other. The inner containment is thus more easily handled, and inspection of the containment and insulating gap is easier to perform. Furthermore, problems connected with thermal expansion of the inner containment are reduced, such as attachment of the container to the pressure vessel and the arrangement of pipes for the supply of feed water to the reactor and for the removal of steam or hot water from the reactor. The nuclear reactor according to the invention is characterized in that the containment consists of an upper part and a lower part, the adjacent parts of the upper part and the lower part being arranged to form a water-filled passage having a U-shaped cross section between the insulating gap and the inside of the reactor, and adjacent parts of the lower part and the pressure vessel being arranged to form a gas-filled lock having an inverted U-shaped cross section between the insulating gap and the inside of the reactor. |
claims | 1. An apparatus for securing an inlet riser to a means for bracing the riser, the apparatus comprising:a saddle shaped to mate with a surface of the inlet riser, wherein the saddle includes a plurality of connection structures;a plurality of riser clamp brackets configured to secure to the means for bracing the riser, wherein,each of the riser clamp brackets includes a mating structure configured to mate with a corresponding connection structure of the plurality of connection structures,a mating of the mating structures and the corresponding connection structures draws the saddle to the riser clamp brackets in only a first direction perpendicular to a longitudinal axis of the inlet riser, andthe mating draws the saddle to the surface of the inlet riser in the first direction. 2. The apparatus of claim 1, wherein the saddle has a semicircular shape matching a portion of an outer perimeter of the inlet riser. 3. The apparatus of claim 1, wherein the connection structures each include a flat surface with a hole, and wherein the mating structures each include a mating post shaped to pass through the hole of the corresponding connection structure. 4. The apparatus of claim 3, wherein the mating post extends in the first direction. 5. The apparatus of claim 4, further comprising:a plurality of fasteners each configured to join to a corresponding mating post of the mating posts after passing through the hole of the corresponding connection structure, wherein the fasteners bias the saddle against the riser clamp brackets in the first direction and bias the saddle against the inlet riser in the first direction. 6. The apparatus of claim 5, wherein the mating posts are threaded and wherein the fasteners are nuts each configured to screw down on the corresponding mating posts in the first direction. 7. The apparatus of claim 1, wherein the saddle is a single, continuous piece that directly mates with the riser clamp brackets and with the inlet riser. 8. The apparatus of claim 1, wherein the saddle is shaped to fit in an axial separation between two side members of the means for bracing the riser and join to the riser clamp brackets in the axial separation. 9. The apparatus of claim 1, wherein each of the riser clamp brackets is a single, continuous piece that is configured to directly secure to the means for bracing the riser. 10. A damping system for use in a nuclear power plant, the system comprising:a reactor pressure vessel;a jet pump assembly including an inlet riser having a longitudinal axis extending vertically parallel to a surface of the reactor pressure vessel;a riser brace extending perpendicular to the longitudinal axis from an attachment wall, wherein,the riser brace includes a plurality of first side members and a plurality of second side members, whereinthe first side members and second side member extend around the inlet riser on opposite sides of the inlet riser,the first side members are separated from each other in the longitudinal direction, andthe second side members are separated from each other in the longitudinal direction, andthe riser brace includes a yoke mating with the first and the second side members and biasing the inlet riser in only a first direction perpendicular to the longitudinal axis and toward the attachment wall;a riser brace clamp including,a saddle shaped to mate with a surface of the inlet riser, wherein the saddle includes a plurality of connection structures,a plurality of riser clamp brackets configured to secure to the riser brace, wherein,each of the riser clamp brackets includes a mating structure configured to mate with a corresponding connection structure of the plurality of connection structures,a mating of the mating structures and the corresponding connection structures draws the saddle to the riser clamp brackets in a second direction opposite the first direction, andthe mating draws the saddle to the surface of the inlet riser in the second direction. 11. The system of claim 10, wherein the saddle has a semicircular shape matching a portion of an outer perimeter of the inlet riser. 12. The system of claim 10, wherein the connection structures each include a flat surface with a hole, and wherein the mating structures each include a mating post shaped to pass through the hole of the corresponding connection structure. 13. The system of claim 12, wherein the mating post extends in the second direction. 14. The system of claim 13, wherein the riser brace clamp further includes a plurality of fasteners each configured to join to a corresponding mating post of the mating posts after passing through the hole of the corresponding connection structure, wherein the fasteners bias the saddle against the riser clamp brackets in the second direction and bias the saddle against the inlet riser in the second direction. 15. The system of claim 14, wherein the mating posts are threaded and wherein the fasteners are nuts each configured to screw down on the corresponding mating posts in the second direction. 16. The system of claim 10, wherein the saddle is a single, continuous piece that directly mates with the riser clamp brackets and with the inlet riser. 17. The system of claim 10, wherein the saddle is shaped to fit and join to the riser clamp brackets between two axially adjacent first side members, and wherein the saddle is shaped to fit and join to the riser clamp brackets between two axially adjacent second side members. 18. The system of claim 10, wherein each of the riser clamp brackets is a single, continuous piece that directly secures to the yoke of the riser brace. 19. The system of claim 10, wherein, for each of the jet pump assembly, there is only a single of the riser brace clamp having only a single of the saddle and only two of the riser clamp brackets. 20. The system of claim 19, wherein the single saddle is shaped to fit and join to the two riser clamp brackets between two axially adjacent first side members, and wherein the single saddle is shaped to fit and join to the two riser clamp brackets between two axially adjacent second side members. |
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