patent_number
stringlengths 0
9
| section
stringclasses 4
values | raw_text
stringlengths 0
954k
|
|---|---|---|
059828382 | description | DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENT The method and apparatus for a neutron based portable system for detection of specific elements is described herein. There are three separate elements which comprise the method and apparatus of the present invention. They include: (i) the portable probe and control equipment; (ii) data de-convolution; and (iii) object identification. Each will be dealt with individually herein. I. Probe Construction The apparatus is shown schematically in FIG. 1 wherein each of the separate elements of the portable apparatus 10 are shown. The portable pulsed neutron detection system 10 is comprised of multiple elements including the probe 10, low voltage power supply 32, portable computer 30, data and power lines 25 and 26. The entire apparatus is lightweight and easily movable which allows the apparatus 10 to be used in many different environments without the need of a heavy, permanently placed stand alone apparatus. The probe 20 is constructed of multiple elements. The probe 20 is encased in a stainless steel cylindrical housing 27 and has a diameter of less than 5 cm. The probe has a length of around 2 meters and weighs around 36-45 kg. Inside the housing 27 is found the electronics and other apparatus for emitting, detecting and controlling the probe. At the distal end of the probe housing are found gamma ray detectors 21 for detection of the gamma rays emitted from the interrogated object after subjecting it to pulsed neutrons. Next in line within the probe 20 is found shielding material 22 which separates the neutron emitter 23 from the detectors 21. The pulsed neutron generator is found adjacent to the shielding material and emits pulses of neutrons at 14 MeV. Finally there is found within the probe housing 27 the electronics which generate the low and high voltage power supplies and electronics which control the gamma ray detectors. The gamma ray detectors 21 found at the distal end of the probe 20 are high efficiency detectors made of Bismuth Germanate or Gadolinium Orthosilicate. The gamma ray detectors 21 are organized in linear geometry. The materials which constitute these gamma ray detectors are important in the operation of the present apparatus and should therefore be able to detect low intensities of the gamma rays. The orientation of the actual detectors within the probe does not particularly affect the efficiency of detection of levels corresponding to carbon, nitrogen and oxygen and a linear orientation is set forth herein for explanatory purposes. Many different orientations are possible as long as the probe calibration is adjusted appropriately. These detectors are temperature stabilized to avoid any changes in the light collection efficiency. Thus, the analog amplitude of the signal generated from the gamma ray detectors 21 does not deviate for a given gamma ray detected. The temperature stabilization is accomplished through a thermo-electrically controlled jacket which surrounds each detector. Each detector is connected to a photomultiplier which changes the optical or light output of the detector to a voltage signal. A resistor chain and preamplifier provide high voltage to the photomultiplier and shape and pre-amplify the analog photomultiplier output. The resistor chain is connected to the base of the photomultiplier tube. After pre-amplification, each gamma ray signal detected by the detectors 21 is sent through the data link 25 to the data acquisition system or computer 30. The data acquisition system 30 is comprised of multiple parts consisting of an amplifier, an analog to digital converter, voltage stabilizers and live-time correction. The live-time correction of the gamma-ray detection signals acts to correct for the delay in which the analog to digital converter is busy analyzing and converting data. Within computer 30, the analog signals provided by the photomultiplier representing the counts of gamma photons detected by detectors 21 during a specified time period are amplified and converted to representative digital values by an analog to digital converter. The amplitudes of the signals received by the computer 30 fluctuate within a range of 0-10 Volts and correspond to the energy channel or level of each gamma ray detected by the gamma ray detectors 21. The analog to digital converter utilized in the data acquisition system converts these amplitude values to digital values for use in the data deconvolution and object identification steps detailed herein. The digital value converted from this amplitude signal represents the number of detected gamma rays. These digital values are then stored in a database on the computer 30. The neutron generator 23 is a sealed tube deuterium tritium neutron generator and provides pulsed neutrons at 14 MeV. Each pulse is several microseconds wide and has a frequency of between 10 kHz and 14 kHz. The pulse separation, depending on the pulse width and pulse frequency, varies between 85 and 90 microseconds. These pulsed neutrons initiate the fast neutron reactions described above within the object being interrogated. Outgoing gamma rays emitted by the object are detected by the gamma ray detectors 21 for a specified time period. The gamma rays are measured and recorded as a single spectrum of data during that time period in the data acquisition system 30. At the end of the neutron pulse, the fast neutrons are thermalized at which point they initiate neutron capture reactions with the object. The gamma rays from these reactions are detected again by the detectors 21 and they are stored similarly as with the prior spectrum data. However, this spectrum data is offset within the computer systems 30 memory so as to have two distinct spectrum data from separate time periods. The ability to measure two separate gamma ray spectra created by these reactions and store said data utilizing the same equipment provides a significant advantage over prior art devices. Previously, detecting and storing two distinct gamma ray spectra has required separate electronic components. Further, the present apparatus and method utilizes data collected from a third spectrum from the activation analysis. To detect elements through activation analysis, the neutron generator bombards the object under interrogation for a few minutes or a few seconds, depending on the material being detected. For example, phosphorus must be bombarded for proper detection for about 5 minutes while sodium only requires about 30 seconds. The generator is then turned off and the activation gamma rays are acquired for a time equal to the bombardment time. The activation data is then further utilized to determine the material contained within the object under interrogation. Radiation shielding 22 and 29 is provided and comprised of a high atomic number material such as lead or bismuth. The shielding 22 separates the gamma ray detectors within the probe from the neutron generator 23. Additional shielding 29 is also provided concentrically around each of the detectors 21, surrounding it through at least 125 degrees, and preferably around 235 degrees. This shielding prevents detection of gamma rays emitted from the generator tube 23 and from materials in the background. The concentric shielding 29 provided on the detectors 21 allows the probe to differentiate the contents of two objects when the probe 20 is placed between them. One spectrum of data may be taken with the concentrical shielding 29 towards one of the objects and another is measured with the shielding towards another object. The preferred embodiment of the probe construction contains shielding about 235 degrees of the detector as is shown in FIG. 8. Voltage supply 24 provides 6 V, 12 V and 24 V low voltage for the detectors electronics and the neutron generator electronics. The voltage supply 24 also provides 900 V to 1300 V high voltage for the detectors 21. Differing values of the supplied voltage may be required due to varying requirements from the specific type of detector utilized. Three thousand volts (3,000 V) are provided for the pulsing of the neutron beam and 110,000 V high voltage is provided for the deuterium acceleration within the neutron generator. The probe 20 can also operate with a high voltage power supply residing external of the cylindrical probe 20 thereby reducing the size of probe 20. Such external connection to a high voltage supply would require a high voltage connector between the probe and supply. Neutron generator controller 32 controls the pulsing of the neutron generator 23 and is designed according to the specific design requirements of the particular application. The controller 32 additionally provides controlling logic for the low voltage supplied by voltage supply 24. Additionally, controller 32 can contain diagnostic electronics for the operation of probe 20. Computer 30 is powered by the low voltage power supply 24 and is connected to the probe via data link 25. Computer 30 conducts the appropriate amplification and conversion of the analog data signals generated by the gamma ray detectors 21. Computer 30 also produces the appropriate offset of the gamma ray spectra so that the same analog to digital converter can be used for the separate acquisition of gamma rays produced from both fast and thermal neutrons on the same detector. The computer 30 additionally contains the appropriate software for data reduction and elemental characterization which allow real time analysis of the data. The relationship of the spectra of data collected by each detector and the manner in which computer 30 stores said data are shown in FIG. 9 wherein the detector 21 is shown as well as memory 40 such that when a signal is provided by the detector 21 through cables 25, the data is typically stored in a single memory space due to the fast detection and storage requirements inherent in these types of detection systems. However, due to the requirement of storing two sets of data generated at two different time periods, said periods being within microseconds of each other, the computer offsets the write portion in memory where the data is stored. The fast neutron signals typically last 14 .mu.s and are represented as signal 1 in FIG. 9 as 48. This signal is stored in the upper portion of the memory storage area for that particular detector 21. After 14 .mu.s, the thermal neutron signal 49 is generated. Without the offsetting of the write portion of the computer for saving these data, the second set of data would overwrite the first set of data detected representing fast neutrons. Thus, in this short time period, computer 30 offsets the write pointers in order to properly store the data for multiple sets of signals detected by detectors 21. This cycle is repeated for several seconds or minutes, depending on the investigation carried out. For the neutron activation data, since no cycling is involved and the neutron generator is turned off, the output from detector 21 in FIG. 9 is stored as signal 3 in a separate part of the computer memory. Upon completion of the data acquisition, the acquired files are transferred to the de-convolution computer code residing within computer 30 for analysis. Data link cables 25 and 26 are coaxial shielded cables that provide the probe with the necessary voltages and carry to the controller 32 and computer 30 the detector signals and diagnostic signals. The cables can have a length of up to 17 m allowing the operation of the probe from varying distances. II. Data De-Convolution Shown in FIGS. 2, 3 and 4 are gamma ray spectra from fast neutron, thermal neutron and neutron activation reactions. Each spectrum contains several gamma-rays produced from chemical elements contained in the interrogated object. The peaks in the spectrum, of FIGS. 2-6, represent particular elements present in the sample while the numbers associated with the peaks represent the energy, in kiloelectron volts, of the respective gamma rays. Data is collected for the specific probe geometry or detector configuration utilized in probe 20. Each probe is required to be properly calibrated based upon actual physical configuration of detectors 21, size of detectors, shielding geometry and position of the generator relative to the detectors. For a given detector and a given detection geometry, geometry meaning the probe configuration, each chemical element produces a characteristic gamma ray spectrum which is called a response spectrum. FIG. 5 shows a response spectrum produced from a carbon sample placed in front of the gamma ray detector bombarded with fast neutrons from the neutron generator 23. A gamma ray spectrum from any innocuous material or from a suspect drug or explosive will contain several chemical elements including hydrogen, carbon, nitrogen and oxygen. Depending on the packaging and surrounding materials, it can also contain elements such as silicon, chlorine, iron, lead or other elements. In the absence of any sample placed in front of the detector, the detector 21 will record gamma rays emanating from the materials surrounding the detector as well as from the materials inside and around the neutron generator 23. This spectrum is called the background spectrum. The de-convolution computer code used for the reduction of the data represents the counts or incidences of gamma rays in each energy channel of the spectrum by the equation: EQU f[i]=.SIGMA..sub.k c.sub.k *m.sub.k [i]+.alpha.*bg[i] (1) where f[i] is the number of counts in the i-th energy channel of the fitted spectrum, c.sub.k is the multiplication coefficient for the response spectrum of the k-th element, m.sub.k [i] is the number of counts in the i-th energy channel of the response spectrum of the k-th element, .alpha. is the multiplication coefficient of the background, and bg[i] is the number of counts in the i-th energy channel of the background. The m.sub.k [i]'s are determined by measuring the spectrum of a sample containing only one chemical element (the response spectrum). The coefficients c.sub.k and .alpha. are determined by the least squares method, minimizing the general x.sup.2 expression: EQU X.sup.2 =.SIGMA..sub.i (y.sub.1 -f[i]).sup.2 .sigma..sub.i.sup.2(2) where y.sub.i and .sigma..sub.i are the measured counts in the i-th energy channel and the statistical error respectively. For a given probe configuration (position of detectors 21 relative to the neutron generator, size of each detector, which as previously indicated are cylindrical and 2.5 to 5 cm in diameter and 5 cm in length, collimation of each detector, etc.), response functions are measured for all the major and minor chemical elements (hydrogen, oxygen, chlorine, etc.) that are expected to be detected. To measure a response function, a sample containing primarily the chemical element of interest is placed in front of the detector. The neutron generator is turned on and a spectrum is accumulated. For some elements, such as chlorine, the response spectrum is due primarily to thermal energy neutrons. In this case the response spectrum will be used for the analysis of data from this particular neutron energy. In other cases such as carbon, a gamma ray response spectrum can be formed from both fast neutron and thermal neutron reactions. Finally, the gamma ray response from thermal, fast neutron and activation reactions may all be utilized. All these response spectra are stored as a library in the computer. As long as the probe configuration is not altered, these spectra can be used irrespective of the locality where the probe is used and the type of object being interrogated. When the probe is to be used for specific interrogation, the probe is placed at the position where the measurement is to take place. Without placing in front of the detector the object to be interrogated, the probe is turned on and a spectrum is taken under the exact same conditions (pulsing frequency, time interval of data accumulation) as it would be used for the measurement of the object under interrogation. In the case of an object that cannot be removed, a spectrum is taken at a distance from the object but in its vicinity. This spectrum is called the background spectrum. Since separate spectra are accumulated for fast neutron and thermal neutron induced reactions, these are two distinct background spectra recorded. While the response spectra remain the same for a given probe configuration, irrespective of where the probe is used, the background spectrum is specific to the measurement. Prior to the initiation of the measurement, the background spectrum is examined, the major and minor chemical elements present in it are recognized and recorded. When a spectrum is accumulated, it is displayed and all major and minor chemical elements which were not observed in the background spectrum are also recorded. The de-convolution software is started and the major and minor chemical elements present in the background and the actual spectrum are listed. The software uses this list of elements for accessing from the library the response spectra of the elements of interest. The computer 30 additionally displays the specific background spectra. The data de-convolution software, using the information generated, utilizes equations (1) and (2) and provides a best fit to the data. To generate the best fit data shown in FIG. 6, multiple spectra are produced based upon assumptions in the content of the object being examined. The de-convolution code assumes certain amounts of each element of interest in the sample being interrogated. These assumed amounts are then utilized to generate a fitted spectrum utilizing the above referenced equations. Multiple iterations of adjusting the amounts of the assumed elements are then attempted by the de-convolution code until a best fit of the experimental spectrum is produced. This process is represented in the schematic of FIG. 10. The results of the best fit can be displayed as shown in FIG. 6. The figure contains the experimental spectrum, the fitted spectrum and the background spectrum. Below the spectra the difference at each energy channel between the experimental and fitted spectrum is displayed. The two horizontal lines above and below the difference spectrum are the 3.sigma. lines indicating 99% confidence limit. A table is also provided with the number of counts (in the form of counts per second or any other time unit desired) for each major and minor element detected above the background, along with the error for this measurement. III. Object Identification The de-convolution software provides information on the chemical elements contained in the interrogated object. This information is utilized in several ways depending on the environment of the object under interrogation. If the investigation is of same size objects under standard geometry conditions, the probe 20 can be calibrated absolutely so that the gamma ray counts for each element would correspond to a specific elemental content. This interrogation would correspond to a fixed probe configuration for interrogation of same-sized objects placed at a fixed position relative to the probe. To calibrate the probe for each element, samples with specific elemental composition are analyzed, and measured gamma ray counts versus elemental concentrations are established. An example of such analysis would be the examination of projectiles of similar size containing the same amount of explosives of differing composition. If the probe cannot be calibrated absolutely because the contents of the same sized objects vary in composition and density, the elemental composition of each object with differing content is established under the same geometry conditions. To differentiate and identify each object, a decision tree is made that depends on the elemental content of certain elements as well as elemental ratios of other elements. Such a decision tree is shown in FIG. 7. When interrogating objects under random conditions, if the elemental content cannot be uniquely determined from the number of counts for each element, elemental ratios can be uniquely determined. Ratios such as carbon to oxygen and carbon to nitrogen can be measured irrespective of the size of the object and position of the probe relative to the object. As an example, drugs hidden within innocuous materials such as coffee or sugar can be found by placing the probe close to the material and measuring the carbon to oxygen ratio. For sugar, the carbon to oxygen ratio is 1.1, for coffee it is 1.8, while for heroin and cocaine, it is 4.2. If the calculated number differs from the ratio expected for coffee or sugar by an amount larger than 3.sigma., one can say with 99% confidence that the material contains something different than coffee or sugar. The foregoing detailed description is given primarily for clearness of understanding and no unnecessary limitations are to be understood therefrom for modifications will become obvious to those skilled in the art upon reading this disclosure and may be made without departing from the spirit of the invention or the scope of the appended claims. |
description | This application claims priority to French application FR 09 50883, filed on Feb. 12, 2009, the entire disclosure of which is incorporated by reference herein. The present invention relates to a method for selecting a loading map for a nuclear reactor core. The invention is used, for example, in pressurized water nuclear reactors. In conventional manner, the core of such a reactor is loaded with more than a hundred juxtaposed nuclear fuel assemblies. Each assembly comprises a bundle of nuclear fuel rods, the rods each having a cladding which surrounds nuclear fuel pellets. The operation of a nuclear reactor is carried out in the form of cycles, at the end of which the nuclear fuel assemblies of the core are replaced, for example, in thirds. In this manner, at the end of an operating cycle, for example, one third of the nuclear fuel assemblies present in the core are removed, two thirds of the nuclear fuel assemblies in the core are moved and a new third of nuclear fuel assemblies are introduced so that they are subjected to the following operating cycle. Conventionally, the loading map, that is to say, the arrangement of the nuclear fuel assemblies within the core, is selected from a plurality of loading maps envisaged in order to produce the maximum amount of energy during the future operating cycle. This selection is brought about, for example, by the operator of the nuclear reactor, using computer programs which involve neutron calculation software. By way of example, documents EP-786 782 and U.S. Pat. No. 7,224,761 describe methods for neutron optimization of loading maps. After determining a loading map which is acceptable on the basis of neutron criteria, it must be verified that, the loading map envisaged does not involve any risk of exceeding the long-term thermomechanical limits within the nuclear fuel assemblies. Exceeding such limits in this manner could lead in particular to occurrences of thermal instability within the core. This verification is generally provided by the designer of the fuel assembly taking into account a notional rod which would be subjected to a local power level greater than all the local power levels of the rods, regardless of the time of the cycle in question; this is referred to as an envelop power history. The duration of such a verification is in the order of from two to three days. After the loading map selected on the basis of neutron criteria is confirmed with respect to the thermomechanical criteria, the operator of the nuclear power station produces a report showing compliance with the safety parameters of the reactor so that the safety authorities definitively validate the loading map selected. Use of nuclear reactors is increasingly demanding, in particular with increases in the duration of operating cycles. It is further desirable to be able to best use the capacities of the reactors and the nuclear fuel assemblies which are loaded therein. An object of the invention is to overcome this problem, by providing a method for selecting a loading map of a nuclear reactor core which allows the nuclear fuel assemblies to be better used whilst ensuring safe operation. The invention provides a method for selecting a loading map for a nuclear reactor core from a plurality of envisaged loading maps, the core comprising nuclear fuel assemblies, each assembly comprising nuclear fuel rods where nuclear fuel is enclosed in a cladding, the method comprising the following steps: a) providing production data relating to the nuclear fuel assemblies, b) providing neutron data which are representative of the past operation of the core of the nuclear reactor and the future operation thereof at least during the next operating cycle, c) calculating, using an item of neutron calculation software and based on the production data and neutron data, the three-dimensional distribution of the local power in the core during future operation for each loading map envisaged, d) calculating, using an item of thermomechanical calculation software and based on the calculated distribution of local power, the extreme value reached by at least one thermomechanical parameter within the nuclear fuel assemblies for each loading map envisaged, e) selecting, in accordance with the extreme values calculated, a loading map from the loading maps envisaged. According to specific embodiments, the method may comprise one or more of the following features, taken in isolation or according to any technically possible combination: step d comprises the following sub-steps: d1) identifying at least one rod for which the value of the thermomechanical parameter is capable of being the extreme value from the rods of the nuclear fuel assemblies; d2) calculating using the thermomechanical calculation software the extreme values reached by the thermomechanical parameter for the rod(s) identified; step d1) is carried out using an item of sorting software; the sorting software uses interpolations or correlations established from the thermomechanical calculation software; the thermomechanical parameter is representative of the stress state in a cladding; the thermomechanical parameter is the pressure inside the cladding and the extreme value calculated is the maximum value of the pressure; the thermomechanical parameter is representative of the corrosion state of a cladding; the thermomechanical parameter is a thickness of oxide on the cladding and the extreme value calculated is the maximum value of the thickness of oxide; steps c) and d) are carried out in a time period of less than one week for each loading map envisaged; steps c) and d) are executed in a time period of less than one day for each loading map envisaged; and a step f) for loading the core in accordance with the loading map selected. The invention also provides a system for selecting a loading map of a nuclear reactor core, characterized in that it comprises means for implementing the steps of a method as defined above. According to one embodiment of the invention, the system comprises at least one data-processing unit and storage means in which there is/are stored at least one programme for carrying out steps of the selection method used by the system. The invention further provides a computer program which comprises instructions for carrying out the steps of a method as defined above. The invention also provides a storage medium which can be used in a computer and on which a program as defined above is recorded. FIG. 1 is a schematic illustration of a pressurized water nuclear reactor core 1, it being understood that the invention can apply to other types of reactor. It is possible to see the vessel 3 of the nuclear reactor, a thermal screen 5 and the barrel 7 of the core 1 inside which nuclear fuel assemblies 16 are arranged. The assemblies 16 are orientated vertically and juxtaposed. They are indicated with squares in FIG. 1. Typically, the core 1 comprises more than a hundred fuel assemblies 16, for example, 157 for a reactor of 900 MWe. As illustrated in FIG. 2, each assembly 16′comprises in conventional manner a network of nuclear fuel rods 24 and a skeleton 26 for supporting the rods 24. The skeleton 26 conventionally comprises a bottom nozzle 28, an upper nozzle 30, guide tubes 31 which connect the two nozzles 28 and 30 and which are intended to receive rods of the control clusters of the reactor and spacer grids 32. As illustrated in FIG. 3, each fuel rod 24 comprises in conventional manner, a cladding 33 in the form of a tube which is closed at the lower end thereof by a bottom end plug 34 and at the upper end thereof by a top end plug 35. The rod 24 comprises a series of pellets 36 which are stacked in the cladding 33 and are in abutment against the bottom end plug 34. A retention spring 38 is arranged in the upper portion of the cladding 33 in order to abut against the upper end plug 35 and the upper pellet 36. Conventionally, the pellets 36 are based on uranium oxide and the cladding 33 is of zirconium alloy. FIG. 4 illustrates the various steps of a method for selecting a loading map for the nuclear fuel assemblies 16 within the core 1 in order to implement a future operating cycle of the reactor. This selection is carried out from a plurality of loading maps envisaged. A loading map comprises in particular the positions which the various nuclear fuel assemblies 16 must occupy and therefore indicates how to move the nuclear fuel assemblies 16 which are already present during the preceding operating cycle and which have to carry out one or more additional operating cycles and/or load the new nuclear fuel assemblies 16 in accordance with their characteristics. The method for selecting the loading map is implemented using a data-processing system 40, which is illustrated in FIG. 5 and which comprises, for example, a data-processing unit 42 which comprises one or more processor(s), data storage means 44, input/output means 46 and optionally display means 48. The storage means 44 which comprise, for example, one or more memory/ies, store one or more computer program(s) which involve various items of software in order to carry out the steps described below. The loading maps envisaged have, for example, been stored therein. In step 102, production data are provided relating to the nuclear fuel assemblies 16 which will be present in the future operating cycle of the core 1. These production data typically comprise, for each assembly 16, information relating to: the dimensions of the various elements which constitute the assembly 16, the material(s) used, for example, to constitute the claddings 33 of the rods 24, the nuclear fuel(s) that the rods 24 contain, in particular in terms of the nature of the fissile material, enrichment and optionally neutron poisons. These production data are, for example, stored in the storage means 44. In step 104, neutron data are provided which are representative of the past operation of the core 1 of the nuclear reactor and the future operation envisaged, in particular during the future operating cycle. The data relating to the past operation can be real data, for example, measured during previous operating cycles using different sensors present in the nuclear reactor. These may also be notional power records simulated by an item of neutron calculation software such as that used in the following steps of the selection method of the loading map. The neutron data are, for example, stored in the storage means 44. During step 106, for an envisaged loading map, the system 40 will calculate, using an item of neutron calculation software, and using the production data and the neutron data provided in steps 102 and 104, the three-dimensional distribution of the local power in the core 1 during the future operation envisaged. The item of neutron calculation software used, may be an item of software of conventional type, for example, the software SMART from the company AREVA NP (registered trade mark). This neutron calculation software is, for example, stored in the storage means 44 of the system 40. Then in step 108, the system 40 will calculate, using an item of thermomechanical calculation software, from the distribution of the local power calculated in step 106 and the production data provided in step 102, the internal pressure Pint inside the claddings 33 of the rods 24 of all the assemblies 16 which will be present in the core 1 during the future operating cycle. The calculation of step 108 can be carried out in two sub-steps. In a first sub-step, the system 40 will identify the rod(s) 24 which is/are the most loaded in each of the assemblies 16 of the core 1 during the future operating cycle, that is to say, the rod(s) 24 of each assembly 16 which are subjected to the greatest internal pressure Pint. This identification can be carried out using an item of sorting software which, rather than carrying out explicit calculations in order to solve the different applicable equations, will determine the Pint values for the different rods 24 by means of correlation from the known Pint values for known conditions (linear power, combustion rate, etc.). The known Pint values originate, for example, from a database which is constructed from an item of thermomechanical calculation software and stored in the storage means 44. This may be a conventional item of software, such as the software COPERNIC from the company AREVA NP (registered trade mark). The sorting software used is preferably a simplified version of the same item of thermomechanical calculation software as that used to construct the database and for step 108, since the use of software based on the same models allows the strength and the reliability of the method to be increased. The use of correlations, rather than explicit calculations, allows the rod(s) 24 which will be the most loaded to be identified in a shorter period of time. In a variant, it is possible to use interpolations rather than correlations. Then, in a second sub-step of step 108, the system 40 carries out the complete thermomechanical calculations on the rod(s) 24 identified during the first sub-step of step 108. These calculations are carried out using an item of thermomechanical calculation software of conventional type, for example, the software COPERNIC from the company AREVA NP. These complete thermomechanical calculations carried out on the various rods 24 identified allow the maximum Pint value to be determined within the core 1 during the future operating cycle. In a variant, step 108 can be carried out on only a portion of the assemblies 16 of the core, for example, only on those having already been irradiated during a previous operating cycle. The maximum Pint values will normally be reached within such assemblies 16 rather than in new assemblies 16. As illustrated by the arrow 110, the steps 106 and 108 are repeated for all of the loading maps envisaged for the future operating cycle. In a variant, it is possible to calculate the power distributions in the core 1 for a plurality of loading maps envisaged during the same step 106, then to calculate the maximum corresponding Pint values for each of these loading maps during step 108. Then, in a step 112, a loading map will be selected from the different loading maps envisaged. This selection is carried out in accordance with the maximum Pint values calculated for each loading map envisaged. In this manner, in this step 112, the maximum Pint values calculated are compared with the admissible limit value Plim for Pint. The loading maps which would cause this limit value Plim to be exceeded are eliminated. From the remaining loading maps, it is possible to select, for example, the one which is closest to the limit value Plim, which will allow best use of the nuclear fuel assemblies 16. It is also possible to select a loading map which places less stress upon the claddings 33, in order to retain a greater safety margin. The selection can also be carried out using other criteria, for example, of the neutron type. Then, in step 114, the core 1 of the nuclear reactor is loaded in accordance with the loading map selected in step 112. It can be seen that this last step will normally be implemented after validation by the relevant safety authorities of the loading map selected. The validation is carried out on the basis of a report which indicates that the loading map proposed allows the safety parameters of the reactor to be complied with for the entire duration of the future operating cycle. The selection of the loading map therefore allows not only neutron criteria to be taken into account, as in the prior art, but also thermomechanical criteria. Therefore, this selection can be optimized with respect to long-term thermomechanical limits which are fixed for the nuclear fuel assemblies 16. Since the thermomechanical calculations are not carried out on a notional rod 24, as in the prior art, the maximum Pint value calculated is found to be much closer to the real maximum value which it envelops without excessive conservatism, so that the assemblies 16 can be much better used. The method described above can be used by the operator of the nuclear reactor to which the necessary items of software have been submitted in their binary excecutable form and are thus stored in the storage means 44 of the system 40 of the operator. The operator therefore no longer needs to turn to the manufacturer of the assemblies 16 in order to carry out thermomechanical calculations. He is capable, in step 112, of carrying out himself the selection of the loading map based on the maximum Pint values determined in step 108 and optionally other data, for example, neutron data. In contrast to a complete calculation of the thermomechanical parameters for all the rods 24 which would take from two to three weeks, the calculations of step 108 are carried out on a limited number of rods and therefore quickly, preferably in less than one week, and even more preferably in less than one day, for each loading map. In this manner, the operator is capable of assessing the relevance of a plurality of loading maps within reasonable periods of time. Some steps of the method above and in particular steps 112 and 114 are not necessarily carried out by the data-processing system 40. Furthermore, other long-term thermomechanical parameters than the pressure Pint can be calculated in step 108. These may be, for example, parameters relating to the state of stress in the claddings 33 other than the pressure Pint. They may also be thermomechanical parameters relating to the corrosion state of the claddings 33, for example, the thickness of oxide. In some cases, the extreme values calculated in step 108 will not be the maximum values, but instead the minimum values. In the same manner, step 108 is not necessarily carried out in two sub-steps. This is in particular the case when the thermomechanical parameter calculated relates to the corrosion state of the claddings 33. In this instance, the step 108 can be carried out by a complete calculation using the thermomechanical calculation software on all the rods 24 of the nuclear fuel assemblies 16 in a prior sub-step for sorting. |
|
claims | 1. An energy reclamation system comprising:at least one thermally conductive canister having a hermetically sealed cavity and a central axis, a basket disposed within the hermetically sealed cavity and comprising a grid of cells containing spent nuclear fuel emanating heat, wherein the grid of cells comprises a first region of cells and a second region of cells circumferentially surrounding the first region of cells, wherein the spent nuclear fuel is arranged so that the spent nuclear fuel contained within the first region of cells has a burnup decay power output greater than the spent nuclear fuel contained within the second region of cells;a storage cavity, the canister disposed within the storage cavity;an air-intake passageway from an ambient environment to a bottom portion of the storage cavity;an air-outlet passageway from a top portion of the storage cavity to an ambient environment; andan energy reclamation unit disposed within the air-outlet passageway. 2. The system of claim 1 wherein the grid of cells comprises a third region of cells circumferentially surrounding the second region of cells, wherein the spent nuclear fuel contained within the second region of cells has a burnup decay power energy output greater than the spent nuclear fuel contained within the third region of cells. 3. The system of claim 2 wherein the first region of cells are centrally located along the axis of the canister. 4. The system of claim 1 wherein the energy reclamation unit is a heat exchanger. 5. The system of claim 1 wherein the basket is constructed of a discontinuously reinforced aluminum/boron carbide metal matrix composite material. 6. The system of claim 5 wherein the canister comprises a canister shell formed of a heat conductive metal. 7. The system of claim 1 wherein the canister is configured to achieve a thermosiphon cyclical flow of gas within the hermetically sealed cavity. 8. An energy reclamation system comprising:at least one thermally conductive canister having a hermetically sealed cavity and a central axis, a basket disposed within the hermetically sealed cavity and comprising a grid of cells, wherein the grid of cells comprises a first region of cells and a second region of cells circumferentially surrounding the first region of cells;a first plurality of spent nuclear fuel rods contained within the first region of cells;a second plurality of spent nuclear fuel rods contained within the second region of cells;wherein the first plurality of spent nuclear fuel contained within the first region of cells has a burnup decay power output greater than the second plurality of spent nuclear fuel contained within the second region of cells;a storage cavity, the canister disposed within the storage cavity;an air-intake passageway from an ambient environment to a bottom portion of the storage cavity;an air-outlet passageway from a top portion of the storage cavity to an ambient environment; andan energy reclamation unit disposed within the air-outlet passageway. 9. An energy reclamation system comprising:at least one thermally conductive canister having a hermetically sealed cavity and a central axis, a basket disposed within the hermetically sealed cavity and comprising a grid of cells, wherein the grid of cells comprises a first region of cells and a second region of cells circumferentially surrounding the first region of cells;a first plurality of spent nuclear fuel rods contained within the first region of cells;a second plurality of spent nuclear fuel rods contained within the second region of cells;wherein the second plurality of spent nuclear fuel contained within the second region of cells has a burnup decay power output less than the second plurality of spent nuclear fuel contained within the second region of cells;a storage cavity, the canister disposed within the storage cavity;an air-intake passageway from an ambient environment to a bottom portion of the storage cavity;an air-outlet passageway from a top portion of the storage cavity to an ambient environment; andan energy reclamation unit disposed within the air-outlet passageway. |
|
description | The present invention relates to steam generators. For example, in an approximately cylindrical-shaped barrel portion of a steam generator in a pressurized-water nuclear power plant, a heat transfer tube group formed of a number of heat transfer tubes that are bent in a U-shape and whose ends are secured to a tube plate is provided in a state where it is covered by a tube-group outer casing. Water (supply water) supplied to an upper portion of an annular channel formed between the inner wall of the barrel portion and the tube-group outer casing flows down through the annular channel, flows into the tube-group outer casing from a lower portion of the annular channel, and rises along the heat transfer tubes. During this process, since pressurized high-temperature coolant from a reactor flows through the heat transfer tubes and heats the heat transfer tubes, the water that is in contact with the outer surfaces of the heat transfer tubes is heated and moves upward while evaporating. By means of a steam/water separator provided above the heat transfer tube group, the water is separated by into steam and hot water, and the hot water is returned to the annular channel whereas the steam is sent, for example, to a turbine in a secondary system after moisture contained in the steam is separated therefrom. Since the supply water takes heat from the high-temperature coolant flowing through the heat transfer tubes, the temperature of the coolant is gradually reduced from an inlet side toward an outlet side. In this way, since the coolant in the heat transfer tubes has a relative temperature difference, an area located from the inlets of the heat transfer tubes to the apexes of the U-shape thereof may also be referred to as a hot side and an area located from the apexes to the outlets thereof may also be referred to as a cold side. In this steam generator, since the supply water is mixed in the entire area of a lower portion of the tube-group outer casing, the temperature of the supply water in this portion is made similar, thus reducing the temperature difference between the coolant in the outlets of the heat transfer tubes and the supply water. Therefore, since the amount of heat exchange is reduced in this portion, the total amount of heat exchange of the steam generator is reduced. Various ways of improving this point have been proposed, and, for example, Patent Citations 1 and 2 describe examples thereof. In the examples, supply water is supplied only to a cold-side portion of an annular channel to reduce the temperature of the supply water flowing into the cold side. This increases the temperature difference between the coolant in the outlets of the heat transfer tubes and the supply water to increase the amount of heat exchange in this portion. Patent Citation 1: Japanese Unexamined Patent Application, Publication No. Hei-03-87501 Patent Citation 2: Japanese Unexamined Patent Application, Publication No. 2000-9888 The hot water separated by the steam/water separator includes fine air bubbles or is likely to generate air bubbles by involving ambient air. Conventionally, water is supplied in a full-scale manner, thereby cooling the hot water to condense such air bubbles. In the technologies described in Patent Citations 1 and 2, since water is not supplied to the hot side, generated air bubbles remain as they are. So-called carry-under, in which the air bubbles descend together with supply water, occurs. When carry-under occurs (is increased), loss of supply water transfer pressure is increased, and thus the supply water does not flow smoothly in the hot side. Accordingly, because the amount of supply water flowing in the hot side is reduced, reducing the amount of heat exchange, the total amount of heat exchange may be reduced. In view of the above-described problems, an object of the present invention is to provide a steam generator in which the total heat transfer efficiency can be improved by suppressing a reduction in the amount of heat exchange at an inlet side of heat transfer tubes while maintaining the effect of increasing the amount of heat exchange at an outlet side thereof. In order to solve the above-described problems, the present invention employs the following solutions. Specifically, according to an aspect, the present invention provides a steam generator including: a heat transfer tube group that is formed of a plurality of heat transfer tubes in each of which both ends thereof are secured to a tube plate and each of which has a U-shaped free end in which a heat medium flows from one end toward the other end; an annular channel that is formed so as to cover the circumference of the heat transfer tube group and that has an opening, at a lower portion thereof, that communicates with the heat transfer tube group; a water supply unit that is disposed at an upper portion of the annular channel and that supplies water to a descending-side portion serving as an area where the heat medium in the heat transfer tubes descends toward the other end; and a steam/water separator that is disposed above the heat transfer tube group and that separates water heated while passing from the annular passage along the circumference of the heat transfer tubes into steam and hot water, in which an air bubble removing member for removing air bubbles is provided in the annular channel, in a rising-side portion serving as an area where the heat medium in the heat transfer tubes rises from the one end. According to the steam generator of this aspect, the water supply unit is disposed at an upper portion of the annular channel and supplies water to a descending-side portion serving as an area where the heat medium in the heat transfer tubes descends toward the other end, in other words, a cold-side portion serving as a relatively-low-temperature area located from the apexes of the heat transfer tubes to the outlets thereof; therefore, it is possible to reduce the temperature of supply water flowing into the descending-side portion, serving as an outlet portion of the heat transfer tubes. When the temperature of the supply water flowing into the descending-side portion is reduced, a large temperature difference can be produced between a relatively-low-temperature heat medium flowing in the outlet portion of the heat transfer tubes and the supply water, and thus the amount of heat exchange can be increased in this portion. Further, since the air bubble removing member is disposed in the annular channel located in the rising-side portion, which serves as an area where the heat medium in the heat transfer tubes rises from the one end, in other words, the hot-side portion, which serves as a relatively-high-temperature area located from the inlets of the heat transfer tubes to the apexes thereof, it is possible to remove air bubbles in the rising-side portion. When air bubbles are removed in an upper portion of the rising-side portion, it is possible to suppress, in the rising-side portion, the occurrence of so-called carry-under, in which air bubbles descend together with supply water. This can suppress an increase in loss of supply water transfer pressure in the rising-side portion, and therefore, it is possible to prevent a reduction in the amount of supply water flowing in the hot side and to maintain the amount of heat exchange in the rising-side portion, as in a conventional manner. Therefore, since a reduction in the amount of heat exchange in the rising-side portion can be suppressed while maintaining the effect of increasing the amount of heat exchange in the descending-side portion of the heat transfer tubes, the total heat transfer efficiency can be improved. In the aspect, a configuration may be used in which the air bubble removing member is a porous plate that has a plurality of through-holes. With this configuration, a flow of supply water stagnates at portions of the porous plate without the through-holes, so that air bubbles that move downward when the supply water descends, in other words, air bubbles that have small volumes and small buoyancies, are accumulated. When the air bubbles are accumulated, they coalesce to form air bubbles that have large volumes. Since the buoyancies of the air bubbles are increased when the volumes thereof are increased, they come up through the through-holes against the descending supply water and are discharged to the space above. In this way, it is possible to suppress, in the rising-side portion, the occurrence of so-called carry-under, in which the air bubbles descend together with supply water. In the configuration, the porous plate may be disposed in the annular channel at a position where the steam/water separator is disposed. In the configuration, the porous plate may be disposed at a position away from the steam/water separator, downward in the annular channel. In the aspect, the air bubble removing member may be a water supply member that supplies water for condensing the air bubbles. In this way, since the water supply member supplies a sufficient amount of water for condensing air bubbles, air bubbles in an upper portion of the rising-side portion are cooled and condensed. Therefore, it is possible to suppress, in the rising-side portion, the occurrence of so-called carry-under, in which the air bubbles descend together with the supply water. Note that the sufficient amount of water for condensing air bubbles, supplied by the water supply member, is, for example, 5 to 10% of the amount of water supplied from the water supply unit, for example, and an influence on the heat-exchange increasing effect in the descending-side portion can be suppressed. In the aspect, in a tube supporting plate that supports the heat transfer tubes, the opening area in the rising-side portion may be larger than the opening area in the descending-side portion. The heat transfer tubes are supported by tube supporting plates that laterally extend at positions located in the vertical direction and that have openings through which supply water passes. According to the present invention, in the tube supporting plates for supporting the heat transfer tubes, the opening areas in the rising-side portion are larger than the opening areas in the descending-side portion, and therefore, the flow resistance in the rising-side portion, that is, the hot-side portion, is reduced. Since supply water flow improves in the rising-side portion when the flow resistance in the rising-side portion is reduced, the supply water easily circulates in the upstream-side portion, and the occurrence of an unstable flow can be suppressed. Therefore, since a reduction in the amount of heat exchange in the rising-side portion can be suppressed, the total heat transfer efficiency can be improved. In the aspect, a heat-transfer-tube partitioning plate that partition the rising-side portion and the descending-side portion may be provided at a lower portion of the heat transfer tube group, and a channel in the descending-side portion partitioned by the heat-transfer-tube partitioning plate may be a serpentine channel. This makes supply water meander upward in the descending-side portion partitioned by the heat-transfer-tube partitioning plate, so that supply water flows in lateral directions. When supply water flows in the lateral directions, the number of chances, in other words, the duration of time, for it to make contact with the heat transfer tubes is increased compared with a case where supply water simply rises, and thus the amount of received heat can be increased accordingly. Therefore, since the heat exchange efficiency in the descending-side portion is improved, the total heat transfer efficiency can be improved. In the aspect, a plurality of openings that make the annular channel communicate with the heat transfer tube group are provided at points with gaps being provided therebetween in a vertical direction, in a lower portion of the heat transfer tube group. Since supply water flows in the lateral direction at the water supply points, the number of chances, in other words, the duration of time, for it to make contact with the heat transfer tubes is increased, and thus the amount of received heat can be increased accordingly. Since water is supplied to the descending-side portion from a plurality of points with gaps being provided therebetween in the vertical direction, the heat exchange efficiency is improved compared with a case where water is supplied from a single lowermost point, and thus the total heat transfer efficiency can be improved. Note that, in this case, it is preferable that the lower portion of the heat transfer tube group be partitioned into the rising-side portion and the descending-side portion, so as to supply water to the descending-side portion. In the aspect, a partitioning plate that vertically divides the annular channel into the rising-side portion and the descending-side portion is further included, in which, in at least a lower portion of the partitioning plate, a circumferential region in the rising-side portion becomes gradually smaller in the downward direction than a circumferential region in the descending-side portion. This supplies water supplied from the water supply unit, also to the rising-side portion, and therefore, the temperature of supply water in a lower position of the rising-side portion can be reduced. When the temperature of the supply water in the lower position of the rising-side portion is reduced, the difference in temperature between heat medium flowing in the heat transfer tubes and the supply water is increased even in the rising-side portion, and therefore, the heat exchange efficiency can be improved. In the aspect, anti-vibration bars inserted between free ends of the heat transfer tubes adjacent to each other may be arranged such that the arrangement density in the rising-side portion is lower than the arrangement density in the descending-side portion. In the aspect, a resistance porous plate having a plurality of through-holes may be further included at a position in the descending-side portion between the heat transfer tube group and the steam/water separator. In this aspect, a plurality of steam/water separators may be provided, of which a steam/water separator that is positioned in the descending-side portion has an orifice channel resistance at an inlet thereof. Alternatively, those structures may be combined. This makes the channel resistance in the descending-side portion larger than that in the rising-side portion, and therefore, supply water flows in the rising-side portion more easily than in the descending-side portion. Accordingly, more supply water flows in the rising-side portion, and therefore, the proportion of air bubbles in the upper portion of the rising-side portion can be reduced. Therefore, since air bubbles can be prevented from being mixed into supply water to be supplied to the rising-side portion in combination with air-bubble removal by an air bubble removing member, it is possible to effectively prevent an unstable flow in the rising-side portion and to improve the heat exchange efficiency. According to the present invention, since the water supply unit that supplies water to the descending-side portion in the upper portion of the annular channel is provided with the air bubble removing member in the rising-side portion, it is possible to increase the amount of heat exchange in the descending-side portion and to maintain the amount of heat exchange in the rising-side portion, as in a conventional manner. As described above, since a reduction in the amount of heat exchange in the rising-side portion can be suppressed while maintaining the effect of increasing the amount of heat exchange in the descending-side portion of the heat transfer tubes, the total heat transfer efficiency can be improved. 1: steam generator 11: tube plate 13: heat transfer tube 15: heat transfer tube group 23: annular channel 25: opening 29: water supply box 31: steam/water separator 35: porous plate 43: through-hole 45: water supply pipe 47: heat-transfer-tube partitioning plate 53: inflow pore 55: partitioning plate 57: resistance porous plate 59: orifice channel resistance D: serpentine channel C: cold-side portion H: hot-side portion Embodiments of the present invention will be described below with reference to the drawings. A first embodiment of the present invention will be described below with reference to FIGS. 1 to 3. In this embodiment, the present invention is applied to a steam generator 1 in a pressurized-water nuclear power plant. Note that this embodiment does not limit the invention. Further, in this embodiment, components that a person skilled in the art can replace and can easily make or components that are substantially the same can be used. FIG. 1 is a longitudinal cross section schematically showing, in outline, the overall structure of the steam generator 1. FIG. 2 is a cross section taken along arrows X-X shown in FIG. 1. FIG. 3 is a partial longitudinal cross section of a porous plate. The steam generator 1 includes a lower barrel 3 having an approximately cylindrical shape and an upper barrel 5 having an approximately cylindrical shape. The lower barrel 3 includes a first water chamber 7, a second water chamber 9, a tube plate 11, a heat transfer tube group 15 formed of a number of heat transfer tubes 13, and a tube-group outer casing 17. The first water chamber 7 and the second water chamber 9 are disposed so as to divide a lower portion of the lower barrel 3 into two. Primary coolant from a reactor is introduced to the first water chamber 7 through a nozzle (not shown). The primary coolant introduced to the first water chamber 7 passes through a plurality of heat transfer tubes 13 and is recovered in the second water chamber 9. The primary coolant recovered in the second water chamber 9 is discharged toward the reactor through a nozzle (not shown). The heat transfer tubes 13 are bent to have a U-shape. Both ends of each of the heat transfer tubes 13 are secured to the tube plate 11 provided on upper portions of the first water chamber 7 and the second water chamber 9, so as to pass through the tube plate 11. The heat transfer tubes 13 extend upward from the tube plate 11, and free ends thereof are positioned at almost an upper end of the lower barrel 3. Attachment positions of both ends of the heat transfer tubes 13 to the tube plate 11 are positions where one end thereof is communicated with the first water chamber 7 and the other end thereof is communicated with the second water chamber 9. The heat transfer tubes 13 are arrayed along planes parallel to each other. In each plane, the heat transfer tubes 13 are arrayed such that the radii of curvature of the free ends thereof are gradually increased from the inner side toward the outer side. The number of included heat transfer tubes 13 is gradually reduced from the top, going outward from the axis of the lower barrel 3. Therefore, the free ends of the heat transfer tubes 13 form an approximately hemispherical shape as a whole, and the lower portions thereof form the heat transfer tube group 15 having an approximately cylindrical shape. A plane connecting the apexes of the respective heat transfer tubes 13 is perpendicular to the plane of the paper of FIG. 1 and extends to pass through the border between the first water chamber 7 and the second water chamber 9. The first water chamber 7 side of this plane is an area in which high-temperature primary coolant introduced to the first water chamber 7 rises through the heat transfer tubes 13, and is therefore referred to as a hot-side portion (rising-side portion) H. On the other hand, the second water chamber 9 side thereof is an area in which the primary coolant whose heat has been exchanged in the hot-side portion and thus whose temperature has been relatively reduced descends through the heat transfer tubes 13 toward the second water chamber 9, and is therefore referred to as a cold-side portion (descending-side portion) C. The tube-group outer casing 17 has an approximately cylindrical shape with one end face thereof being opened and is provided so as to cover the heat transfer tube group 15. Straight tube portions of the heat transfer tubes 13 are supported by a plurality of tube supporting plates 19 provided in a vertical direction with gaps therebetween. Each of the tube supporting plates 19 has through-holes 21 (see FIG. 6) in portions where the heat transfer tubes 13 are supported. A gap is provided between the tube-group outer casing 17 and the inner wall of the lower barrel 3 and the upper barrel 5, so as to form an annular channel 23 through which supply water flows. The tube-group outer casing 17 and the heat transfer tube group 15 are substantially united via the tube supporting plates 19. A gap is provided between a lower end position of the tube-group outer casing 17 and an upper end position of the tube plate 11 so as to form an opening 25 for communicating between the inside (the heat transfer tube group 15) of the tube-group outer casing 17 and the annular channel 23. At the free end portion of the heat transfer tubes 13, anti-vibration bars 27 are inserted between adjacent rows to prevent vibration of the heat transfer tubes 13. The upper barrel 5 includes a water supply box (water supply unit) 29, a plurality of steam/water separators 31, a moisture separator 33, a porous plate (air bubble removing member) 35, and a steam chamber 37. The water supply box 29 has a semi-ring shape and is mounted at a lower portion of the upper barrel 5 along the cold-side portion C at the upper portion of the tube-group outer casing 17. The water supply box 29 supplies the cold-side portion C of the annular channel 23 with secondary-coolant-system water supplied through a supply-water inlet nozzle 39. A plurality of, for example, 20, steam/water separators 31 are provided on the upper surface of the tube-group outer casing 17 with appropriate gaps being provided therebetween. The steam/water separators 31 crudely separate steam mixed with water that has passed through the heat transfer tube group 15 into steam and water (hot water). The water separated by the steam/water separators 31 is returned to the annular channel 23. The steam crudely separated by the steam/water separators 31 is introduced to the moisture separator 33, and moisture contained in the steam is separated therefrom. The steam from which moisture has been separated and eliminated is sent from the steam chamber 37 through a steam outlet nozzle 41 to a turbine in a secondary system, for example. The porous plate 35 is a plate member having an approximately semicircular shape and has a number of through-holes 43 passing therethrough in a thickness direction. The porous plate 35 is mounted above the tube-group outer casing 17 so as to extend laterally while covering the hot-side portion H of the upper barrel 5. As indicated by the two-dot chain line in FIG. 1, the porous plate 35 may be provided at half the height of the annular channel 23, that is, at a position between the tube plate 11 and the steam/water separators 31, so as to cover the hot-side portion H of the tube-group outer casing 17. The operation of the thus-configured steam generator 1 according to this embodiment will be described. High-temperature and high-pressure primary coolant from a reactor (not shown) is introduced to the first water chamber 7 through a nozzle (not shown). This primary coolant is introduced from the first water chamber 7 to one end of each of the heat transfer tubes 13, circulates through the respective heat transfer tubes 13, and is discharged from the other end of each of the heat transfer tubes 13 to the second water chamber 9. The primary coolant is returned to the reactor from the second water chamber. This circulation of the primary coolant keeps the heat transfer tubes 13 at a high temperature. At this time, a temperature gradient is produced by heat exchange with secondary-coolant-system water such that the temperature of the heat transfer tubes 13 is the highest at the first water chamber 7 side and is gradually reduced toward the second water chamber 9 side. The secondary-coolant-system water supplied from the water supply box 29 is supplied to the cold-side portion C of the annular channel 23. Since this water is mixed with high-temperature hot water flowing out from the steam/water separators 31 at the cold-side portion C and cools this hot water, air bubbles included in the hot water are condensed and almost completely removed. Although this mixed water partially moves to the hot-side portion H, it mainly flows down in the cold-side portion C of the annular channel 23 and is supplied to the inside of the tube-group outer casing 17 through the opening 25. In this way, since the secondary-coolant-system water supplied to the cold-side portion C of the heat transfer tube group 15 has a relatively-low temperature, it is possible to produce a large temperature difference between the heat transfer tubes 13 in the second water chamber 9 side, having a relatively-low temperature, and the secondary-coolant-system water. Therefore, the amount of heat exchange can be increased in this portion. On the other hand, hot water flowing out from the steam/water separators 31 that are disposed in the hot-side portion H is mainly supplied to the hot-side portion H of the annular channel 23. This hot water includes fine air bubbles that have not been separated by the steam/water separators 31 or is likely to generate air bubbles by involving ambient air. Water that mainly includes this hot water moves downward through the through-holes 43 in the porous plate 35. At this time, since the flow of the water stagnates at portions of the porous plate 35 without the through-holes 43 and the like, air bubbles k that move downward when the water descends, in other words, air bubbles k that have small volumes and small buoyancies, are accumulated at these portions. The accumulated air bubbles k collide with each other and coalesce to form air bubbles K that have large volumes. Since the buoyancies of the air bubbles K are increased when the volumes thereof are increased, the air bubbles K come up through the through-holes 43 against the descending water and are discharged to the space above. As described above, it is possible to suppress, in the hot-side portion H, the occurrence of so-called carry-under, in which the air bubbles k descend together with secondary-coolant-system water. Although this water partially moves to the cold-side portion C, it mainly flows down in the hot-side portion H of the annular channel 23 and is supplied to the inside of the tube-group outer casing 17 through the opening 25. Since air bubbles can be removed by the porous plate 35 in this way, an increase in loss of water transfer pressure can be suppressed in the hot-side portion H. Therefore, since a reduction in the amount of supply water flowing in the hot-side portion H can be prevented, the amount of heat exchange in the hot-side portion H of the heat transfer tube group 15 can be maintained, as in a conventional manner. Since a reduction in the amount of heat exchange in the hot-side portion H can be suppressed while maintaining the effect of increasing the amount of heat exchange in the cold-side portion C of the heat transfer tube group 15, as described above, the total heat transfer efficiency can be improved. In this way, part of secondary-coolant-system water evaporates to become steam when it rises around the heat transfer tubes 13, which are maintained at a high temperature. The water mixed with the steam is introduced to the steam/water separators 31 and is crudely separated into steam and water. The steam crudely separated by the steam/water separators 31 is introduced to the moisture separator 33, and moisture contained therein is removed therefrom. The steam from which moisture has been separated by the moisture separator 33 is sent to a turbine in the secondary system (not shown) from the steam outlet nozzle 41 provided on the top of the upper barrel 5. The turbine is rotationally driven by this steam, and its power drives a power generator, for example, to generate electric power. Next, a second embodiment of the present invention will be described using FIGS. 4 to 7. In this embodiment, the configuration of the tube supporting plates 19 is different from that of the first embodiment, and the other components are the same as those in the first embodiment; therefore, the difference therebetween will be mainly described below and a repeated description of the other components will be omitted. Note that identical reference symbols are assigned to the same components as those in the first embodiment, and a description thereof will be omitted. FIG. 4 is a longitudinal cross section schematically showing, in outline, the overall structure of the steam generator 1. FIG. 5 is a cross section taken along arrows Y-Y shown in FIG. 4. In this embodiment, the sizes of the through-holes 21 in the tube supporting plates 19 are different between the hot-side portion H and the cold-side portion C. Specifically, the sizes of the through-holes 21 in the hot-side portion H, shown in FIG. 7, are made larger than those of the through-holes 21 in the cold-side portion C, shown in FIG. 6. Therefore, the opening areas of the tube supporting plates 19 in the hot-side portion H are larger than the opening areas thereof in the cold-side portion C. This makes the flow resistance in the hot-side portion H smaller than that in the cold-side portion C. Since water flow improves in the hot-side portion H when the flow resistance in the hot-side portion H is made smaller, the water easily circulates in the hot-side portion H, and the occurrence of an unstable flow can be suppressed. Therefore, since a reduction in the amount of heat exchange in the hot-side portion H can be suppressed, the total heat transfer efficiency of the steam generator 1 can be improved. Next, a third embodiment of the present invention will be described using FIGS. 8 and 9. In this embodiment, the basic configuration is the same as that of the first embodiment, whereas the configurations of the air bubble removing member and the cold-side portion C of the heat transfer tube group 15 are different. The differences therebetween will be mainly described below, and a repeated description of the other components will be omitted. Note that identical reference symbols are assigned to the same components as those in the first embodiment, and a description thereof will be omitted. FIG. 8 is a longitudinal cross section schematically showing, in outline, the overall structure of the steam generator 1. FIG. 9 is a cross section taken along arrows Z-Z shown in FIG. 8. In this embodiment, as an extension of the water supply box 29, a water supply pipe (water supply member) 45 having a semi-ring shape is positioned in the hot-side portion H of the annular channel 23 and is mounted to supply water to that portion. The water supply pipe 45 is branched to supply water from the water supply box 29. The water supply pipe 45 has a pipe diameter smaller than the water supply box 29 and the amount of water supplied is sufficient to condense air bubbles included in water in the hot-side portion H. This amount is, for example, 5 to 10% of the amount of water the water supply box 29 supplies to the cold-side portion C. As described above, since the water supply pipe 45 supplies a sufficient amount of water for condensing air bubbles, air bubbles in an upper portion of the hot-side portion H of the annular channel 23 are cooled and condensed by this water. Therefore, it is possible to suppress, in the hot-side portion H, the occurrence of so-called carry-under, in which air bubbles descend together with supply water. Although this water partially moves to the cold-side portion C, it mainly flows down in the hot-side portion H of the annular channel 23 and is supplied to the inside of the tube-group outer casing 17 through the opening 25. Since air bubbles can be removed by the cold water supplied by the water supply pipe 45 in this way, an increase in loss of water transfer pressure in the hot-side portion H can be suppressed. Therefore, since a reduction in the amount of supply water flowing in the hot-side portion H can be prevented, the amount of heat exchange in the hot-side portion H of the heat transfer tube group 15 can be maintained, as in a conventional manner. Further, in this embodiment, a heat-transfer-tube partitioning plate 47 that partitions the hot-side portion H and the cold-side portion C is provided at a lower portion of the heat transfer tube group 15. Semicircular-shaped flow adjusting plates 49 are attached to the cold-side portion C of the heat-transfer-tube partitioning plate 47 so as to be aligned with the second and fourth tube supporting plates 19 from the bottom. Semi-ring-shaped flow adjusting plates 51 are attached to the cold-side portion C of the tube-group outer casing 17 so as to be aligned with the first and third tube supporting plates 19 from the bottom. With the flow adjusting plates 49 and the flow adjusting plates 51, a water flow channel in the cold-side portion C serves as a serpentine channel D. In this way, since water meanders upward along the serpentine channel D in the cold-side portion C of the heat transfer tube group 15 partitioned by the heat-transfer-tube partitioning plate 47, the water flows in lateral directions. When the water flows in the lateral directions, the number of chances, in other words, the duration of time, for it to make contact with the heat transfer tubes 13 is increased compared with a case where the water simply rises, and thus the amount of heat received from the heat transfer tubes 13 can be increased accordingly. Therefore, since the heat exchange efficiency in the cold-side portion C is further improved, it is possible to improve the total heat transfer efficiency of the steam generator 1, in combination with the fact that the amount of heat exchange in the hot-side portion H can be maintained, as in a conventional manner. Next, a fourth embodiment of the present invention will be described using FIG. 10. In this embodiment, the basic configuration is the same as that of the first embodiment, whereas the configuration of the water supply box 29 is different. The difference therebetween will be mainly described below, and a repeated description of the other components will be omitted. Note that identical reference symbols are assigned to the same components as those in the first embodiment, and a description thereof will be omitted. FIG. 10 is a longitudinal cross section schematically showing, in outline, the overall structure of the steam generator 1. In this embodiment, a lower portion of the water supply box 29 extends to almost a lower end position of the tube-group outer casing 17. Inflow pores 53 that extend in a lateral direction to make the inside of the tube-group outer casing 17 communicate with the inside of the water supply box 29 are provided in the lower portion of the water supply box 29 at positions between the first and second tube supporting plates 19 from the bottom, between the second and third tube supporting plates 19 from the bottom, and between the third and fourth tube supporting plates 19 from the bottom. An appropriate number of inflow pores 53 may be provided as needed. The heat-transfer-tube partitioning plate 47, which partitions the hot-side portion H and the cold-side portion C, is provided at the lower portion of the heat transfer tube group 15. The top of the heat-transfer-tube partitioning plate 47 is set higher than the position of the top inflow pore 53. Secondary-coolant-system water supplied from the water supply box 29 passes through the opening 25 and the respective inflow pores 53 and is supplied in the lateral direction to the cold-side portion C inside the tube-group outer casing 17. In other words, since water flows in the lateral direction, the number of chances, in other words, the duration of time, for it to make contact with the heat transfer tubes 13 is increased. Since the amount of received heat can be accordingly increased, the heat exchange efficiency can be improved. Furthermore, since water is supplied through a plurality of points with gaps being provided therebetween in the vertical direction, the heat exchange efficiency is improved compared with a case where water is supplied through a single lowermost point. Therefore, the total heat transfer efficiency is improved. Next, a fifth embodiment of the present invention will be described using FIGS. 11 to 14. In this embodiment, the basic configuration is the same as that of the first embodiment, whereas the configuration of the annular channel 23 is different. The difference therebetween will be mainly described below, and a repeated description of the other components will be omitted. Note that identical reference symbols are assigned to the same components as those in the first embodiment, and a description thereof will be omitted. FIG. 11 is a longitudinal cross section schematically showing, in outline, the overall structure of the steam generator 1. FIG. 12 is a cross section taken along arrows U-U shown in FIG. 11. FIG. 13 is a cross section taken along arrows V-V shown in FIG. 11. FIG. 14 is a cross section taken along arrows W-W shown in FIG. 11. In this embodiment, two partitioning plates 55 that extend vertically and that partition the annular channel 23 laterally are attached to the annular channel 23. Upper portions of the partitioning plates 55 are positioned at the border between the hot-side portion H and the cold-side portion C to divide the annular channel 23 into the hot-side portion H and cold-side portion C. Lower portions of the partitioning plates 55 are gradually curved toward the hot side in the downward direction. Therefore, a hot-side circumferential region 23H of the hot-side portion H of the annular channel 23 partitioned by the partitioning plates 55 is made gradually smaller than a cold-side circumferential region 23C of the cold-side portion C, as shown in FIGS. 13 and 14. Secondary-coolant-system water is supplied from the water supply box 29 to the cold-side circumferential region 23C of the annular channel 23 and flows down. The cold-side circumferential region 23C enlarges downward so as to enter the area of the hot-side portion H. Therefore, water entering the inside of the tube-group outer casing 17 through the opening 25 positioned at the hot-side portion H is a mix of relatively-hot water that has passed through the hot-side circumferential region 23H and relatively-cold water that has passed through the cold-side circumferential region 23C, thereby reducing the temperature of water to be supplied to the heat transfer tube group in the hot-side portion H. When the temperature of the supply water at a lower position in the hot-side portion H is reduced, the difference in temperature between the heat transfer tubes 13 and the supply water is increased even in the hot-side portion H, and therefore, it is possible to improve the heat exchange efficiency in the hot-side portion H. Next, a sixth embodiment of the present invention will be described using FIG. 15. In this embodiment, the basic configuration is the same as that of the first embodiment, whereas the configuration of the anti-vibration bars 27 is different. The difference therebetween will be mainly described below, and a repeated description of the other components will be omitted. Note that identical reference symbols are assigned to the same components as those in the first embodiment, and a description thereof will be omitted. FIG. 15 is a schematic view showing an upper portion of the heat transfer tube group 15 in this embodiment. In this embodiment, the anti-vibration bars 27 are arranged such that the arrangement density in the hot-side portion H is lower than that in the cold-side portion C. The anti-vibration bars 27 serve as channel resistance when water flows. In this embodiment, since the arrangement density of the anti-vibration bars 27 in the hot-side portion H is lower than that in the cold-side portion C, the channel resistance in the cold-side portion C is larger than that in the hot-side portion H, and supply water flows in the hot-side portion H more easily than in the cold-side portion C. Accordingly, more supply water flows in the hot-side portion H, and therefore, the proportion of air bubbles in the upper portion of the hot-side portion H can be reduced. Therefore, since air bubbles can be prevented from being mixed into supply water to be supplied to the hot-side portion H in combination with air-bubble removal by an air bubble removing member such as the porous plate 35, it is possible to effectively prevent an unstable flow in the hot-side portion H and to improve the heat exchange efficiency. Next, a seventh embodiment of the present invention will be described using FIG. 16. In this embodiment, the basic configuration is the same as that of the first embodiment, whereas the inner configuration of the tube-group outer casing 17 is different. The difference therebetween will be mainly described below, and a repeated description of the other components will be omitted. Note that identical reference symbols are assigned to the same components as those in the first embodiment, and a description thereof will be omitted. FIG. 16 is a longitudinal cross section schematically showing the upper barrel 5 in this embodiment. In this embodiment, a resistance porous plate 57 having a plurality of through-holes is mounted between the heat transfer tube group 15 and the steam/water separators 31 at a position in the cold-side portion C in the tube-group outer casing 17. The resistance porous plate 57 serves as channel resistance when water flows. Since the resistance porous plate 57 is mounted in the cold-side portion C, the channel resistance in the cold-side portion C is larger than that in the hot-side portion H, and supply water flows in the hot-side portion H more easily than in the cold-side portion C. Accordingly, more supply water flows in the hot-side portion H, and therefore, the proportion of air bubbles in the upper portion of the hot-side portion H can be reduced. Therefore, since air bubbles can be prevented from being mixed into supply water to be supplied to the hot-side portion H in combination with air-bubble removal by an air bubble removing member such as the porous plate 35, it is possible to effectively prevent an unstable flow in the hot-side portion H and to improve the heat exchange efficiency. Next, an eighth embodiment of the present invention will be described using FIG. 17. In this embodiment, the basic configuration is the same as that of the first embodiment, whereas the configurations of the steam/water separators 31 are different. The differences therebetween will be mainly described below, and a repeated description of the other components will be omitted. Note that identical reference symbols are assigned to the same components as those in the first embodiment, and a description thereof will be omitted. FIG. 17 is a longitudinal cross section schematically showing the upper barrel 5 in this embodiment. In this embodiment, orifice channel resistances 59 are mounted at inlet ports of the steam/water separators 31 that are positioned in the cold-side portion C. The orifice channel resistances 59 serve as channel resistance when water flows. Since the orifice channel resistances 59 are mounted to the steam/water separators 31 that are positioned in the cold-side portion C, the channel resistance in the cold-side portion C is larger than that in the hot-side portion H, and supply water flows in the hot-side portion H more easily than in the cold-side portion C. Accordingly, more supply water flows in the hot-side portion H, and therefore, the proportion of air bubbles in the upper portion of the hot-side portion H can be reduced. Therefore, since air bubbles can be prevented from being mixed into supply water to be supplied to the hot-side portion H in combination with air-bubble removal by an air bubble removing member such as the porous plate 35, it is possible to effectively prevent an unstable flow in the hot-side portion H and to improve the heat exchange efficiency. The embodiments of the present invention have been described above in detail with reference to the drawings; however, the specific configurations are not limited to those embodiments, and design changes etc. can be included without departing from the scope of the present invention. For example, the embodiments may be combined with each other and used. |
|
description | This application is a division of U.S. application Ser. No. 13/032,172, filed on Feb. 22, 2011, now U.S. Pat. No. 8,242,474, and claims the benefit of priority from the Japanese Patent Application No. 2010-036046 filed on Feb. 22, 2010 and Japanese Patent Application No. 2010-294239 filed on Dec. 28, 2010; the entire contents of which are incorporated herein by reference. 1. Technical Field The present disclosure relates to an extreme ultraviolet (EUV) light generation apparatus in which a target material is irradiated with a laser beam to thereby generate the EUV light. 2. Related Art In recent years, as semiconductor processes become finer, photolithography has been making rapid progress toward finer fabrication. In the next generation, microfabrication at 70 nm to 45 nm, further, microfabrication at 32 nm and beyond will be required. Accordingly, in order to fulfill the requirement for microfabrication at 32 nm and beyond, for example, an exposure device is expected to be developed where an EUV light generation apparatus for generating EUV light having a wavelength of approximately 13 nm is combined with reduced projection reflective optics. As the EUV light generation apparatus, there are three kinds of light generation apparatuses, which include an LPP (laser produced plasma) type light generation apparatus using plasma generated by irradiating a target material with a laser beam, a DPP (discharge produced plasma) type light generation apparatus using plasma generated by electric discharge, and an SR (synchrotron radiation) type light generation apparatus using orbital radiation. An EUV light generation apparatus in accordance with one aspect of the present disclosure is used in combination with a laser system, and the apparatus may include: a chamber provided with at least one inlet port for introducing a laser beam outputted from the laser system into the chamber; a target supply unit provided to the chamber for supplying a target material to a predetermined region inside the chamber, where the target material is irradiated with the laser beam; at least one optical element disposed inside the chamber; a magnetic field generation unit for generating a magnetic field around the predetermined region; an ion collection unit disposed in a direction of a line of magnetic force of the magnetic field for collection an ion generated when the target material is irradiated with the laser beam and flowing along the line of magnetic force; and a gas introduction unit for introducing an etching gas into the chamber. These and other objects, features, aspects, and advantages of the present disclosure will become apparent to those skilled in the art from the following detailed description, which, taken in conjunction with the annexed drawings, discloses preferred embodiments of the present disclosure. Hereinafter, embodiments for implementing the present disclosure will be described in detail with reference to the accompanying drawings. In the subsequent description, each drawing merely illustrates shape, size, and positional relationship of members schematically to the extent that enables the content of the present disclosure to be understood. Accordingly, the present disclosure is not limited to the shape, the size, and the positional relationship of the members illustrated in each drawing. In order to simplify the drawings, a part of hatching along a section is omitted. Further, numerical values indicated hereafter are merely preferred examples of the present disclosure. Accordingly, the present disclosure is not limited to the indicated numerical values. First, an EUV light generation apparatus in accordance with a first embodiment of the present disclosure will be described in detail with reference to the drawings. FIG. 1 is a sectional view illustrating a schematic configuration of an exposure device and the EUV light generation apparatus in accordance with the first embodiment. As shown in FIG. 1, an EUV light generation apparatus 1 in accordance with the first embodiment includes a well-sealed chamber 10, a droplet generator 16 for supplying a target material in a form of a liquid droplet (droplet D) toward a plasma generation site P1 located at substantially the center of the chamber 10, and a droplet collection unit 17 for collecting a droplet which has passed the plasma generation site P1. The chamber 10 may be provided with an exhaust pump (not shown). The droplet generator 16 stores a liquid-state target material, such as molten tin (Sn), which serves as a material for generating EUV light, and is provided with a nozzle 16a having an opening at its tip facing toward the plasma generation site P1. The liquid-state target material stored in the droplet generator 16 is outputted as a liquid droplet D through the nozzle 16a toward the plasma generation site P1. It is to be noted that the target material is not limited to the molten Sn, and various molten metals or other materials may be employed in accordance with a desired wavelength of EUV light to be obtained. Further, a target is not limited to a liquid target, but a solid target may be used. As shown in FIG. 1, the EUV light generation apparatus 1 includes an EUV collector mirror 12 that selectively reflects EUV light emitted at the plasma generation site P1. A laser beam L1 outputted from a driver laser device (not shown) enters the chamber 10 through a window 11 in synchronized timing as the droplet D arrives at the plasma generation site P1. The laser beam L1, traveling through a through-hole 12a provided at substantially the center of the EUV collector mirror 12 from a rear side (opposite side of the reflective side) of the EUV collector mirror 12, is focused by a focusing optical system (not shown) on the droplet D at the plasma generation site P1. With this, the droplet D is ionized and turned into plasma around the plasma generation site P1, and EUV light L2 is emitted from this plasma. A beam dump 18 for preventing the laser beam L1 from entering into an exposure device 100 may be provided to an opposite side from the EUV collector mirror 12 on an axis AX. The EUV light L2 emitted at the plasma generation site P1 is selectively reflected by the EUV collector mirror 12. When the EUV collector mirror 12, for example, is an ellipsoidal concave mirror having a first focus and a second focus, the EUV collector mirror 12 is disposed such that the first focus coincides with the plasma generation site P1 where the EUV light is emitted and the second focus coincides with an intermediate focus IF where the reflected EUV light is focused. With this configuration, the EUV light emitted at the plasma generation site P1 can be imaged at the intermediate focus IF. The configuration is such that the intermediate focus IF is located inside an exposure device connection unit 19 which serves to connect the EUV light generation apparatus 1 and the exposure device 100. Accordingly, the EUV light L2 generated at the plasma generation site P1, having been selectively reflected by the EUV collector mirror 12, is focused at the intermediate focus IF inside the exposure device connection unit 19. Note that although a beam axis of the laser beam L1 coincides with the axis AX of the EUV light L2 in this embodiment, they do not necessarily have to coincide with each other. The EUV light L2 focused at the intermediate focus IF is thereafter introduced into the exposure device 100. In the exposure device 100, a stage 110 which is movable horizontally and which holds a workpiece W101, such as a semiconductor substrate or a glass substrate, to be processed, and an exposure optical system including one or more mirrors M71 through M75 and a reflective mask M73 which shapes the profile of the inputted EUV light L2 to a pattern to be transferred and focuses the EUV light L2 on the workpiece W101 on the stage 110. Thus, the pattern on the mask M73 by the EUV light L2 having propagated through the exposure optical system is imaged on the workpiece W101, whereby a desired pattern is transferred. More specifically, the exposure device 100 includes an illumination optical system for illuminating the mask M73, and a reduced projection optical system for performing reduced projection of the mask pattern on the mask M73 and images onto the workpiece W101 such as a wafer. The illumination optical system includes the mirrors M71 and M72 and illuminates the reflective mask M73. The reduced projection optical system includes the mirrors M74 and M75 and images the mask pattern on the reflective mask M73 onto the workpiece W101. FIG. 2 is a sectional view illustrating a schematic configuration of the EUV light generation apparatus in accordance with the first embodiment. FIG. 2, however, is a section taken along a plane containing the axis AX and along a plane different from that shown in FIG. 1. As shown in FIG. 2, the EUV light generation apparatus 1 includes, in addition to the configuration shown in FIG. 1, a pair of magnets 14 disposed outside the chamber 10, a pair of ion collection units 13 disposed on a central line of magnetic force lines MF of a magnetic field generated by the pair of magnets 14, and an exhaust pump 20 for maintaining the pressure inside the chamber low. The magnets 14, for example, are superconducting electromagnets constituting a magnetic field generation unit, and a magnetic field in a range of 0 through approximately 2 T (tesla) can be generated steadily in a large space by applying current to coils 14a thereof. The pair of magnets 14 is disposed such that they face each other with the plasma generation site P1 located therebetween and such that they are coaxially aligned with their axes passing though the plasma generation site P1. Thus, a charged particle such as an Sn ion generated around the plasma generation site P1 (hereinafter simply referred as ionized debris) is trapped in the magnetic field generated so as to pass through the plasma generation site P1. The trapped ionized debris is subjected to the Lorentz force from the magnetic field, and travels along the magnetic force lines MF revolving around the magnetic force lines MF. With this, an ion flow FL of which a sectional area along a plane perpendicular to the direction of the magnetic force lines MF is limited within a certain range is formed. The ion flow FL flows in two directions along the direction of the magnetic field away from the plasma generation site P1. Further, a pair of ion collection units 13 is disposed on the central line of the magnetic force lines MF of the magnetic field generated by the pair of magnets 14 at positions where the ion collection units 13 face each other with the plasma generation site P1 located therebetween. The ionized debris generated around the plasma generation site P1 and having been turned into the ion flow FL, being trapped in the magnetic field, flows along the magnetic force lines MF, to thereby be collected into either one of the ion collection units 13. Further, as shown in FIG. 2, the EUV light generation apparatus 1 includes an etching gas introduction unit 15 disposed to face an optical element such as a reflective surface of the EUV collector mirror 12 disposed inside the chamber 10. The etching gas introduction unit introduces an etching gas into the chamber 10 for etching a deposited target material (Sn). The target material deposited on the optical element reacts with the etching gas introduced by the etching gas introduction unit to thereby be removed. As the etching gas, hydrogen (H2) gas, halogen gas such as F, Cl, and Br, or a gas containing the above may be used. When, for example, Sn is used as the target material and hydrogen gas is used as the etching gas, Sn deposited on the optical element reacts with hydrogen to generate gaseous SnH4, to thereby be removed off a surface of the optical element. Note that the gas (such as SnH4) generated through a reaction between the target material and the etching gas may be discharged from the chamber 10 by the exhaust pump 20 or the like connected to the chamber 10. A free radical source for dissociating (turning into a free radical) the etching gas may be provided around an outlet port of the etching gas introduction unit 15. Examples of the free radical source include a plasma source, W-filament, and microwave. With this, the etching gas can be supplied into the chamber 10 as a free radical particle having a high reactivity with the deposited target material. This makes it possible to efficiently remove the target material deposited on the optical element inside the chamber 10. When, for example, hydrogen gas is used as the etching gas and Sn is used as the target material, a hydrogen radical and Sn react efficiently to thereby generate gaseous SnH4. Here, hydrogen gas may be turned into a hydrogen radical with light having a wavelength of ultraviolet or the like emitted at the plasma generation site P1. FIG. 3 is a sectional view illustrating a schematic configuration of the EUV light generation apparatus in accordance with the first embodiment, to which a unit for introducing the etching gas into the chamber and a sensor for detecting pressure inside the chamber are added. The unit for introducing the etching gas includes an etching gas tank 24, a mass flow controller MFC22 for controlling a flow rate of the etching gas, and the etching gas introduction unit 15. Further, the chamber 10 is provided, via piping, with a pressure sensor 25 for measuring the pressure inside the chamber. With an EUV light generation apparatus 1A configured as such, the pressure inside the chamber 10 can be regulated by adjusting the flow rate of the etching gas or a discharge rate of the exhaust pump. FIG. 4 is a graph showing a relationship between a magnetic flux density generated at the plasma generation site P1 and a deposition rate of Sn to be deposited on a surface of the EUV collector mirror 12, in a case where the etching gas is not introduced into the chamber. In FIG. 4, the vertical axis represents the deposition rate of Sn, and the horizontal axis represents the magnetic flux density indicating the strength of the magnetic field. When the magnetic flux density is between 0 T and 0.5 T, the deposition rate of Sn has monotonically decreased. On the other hand, in a region where the magnetic flux density is at or above 0.5 T, the deposition rate of Sn has been substantially constant. Based on this result, it is contemplated that the ionized debris is trapped in the magnetic field when the magnetic flux density is in a range between 0 and 0.5 T. It is also contemplated, however, that neutral particles and minute particles which are not charged and cannot be trapped in the magnetic field are deposited when the magnetic flux density is at or above 0.5 T. Thus, in order to etch the deposited Sn of the uncharged neutral particles and minutes particles, hydrogen gas containing hydrogen radicals has been introduced as the etching gas into the vacuum chamber. In this example, hydrogen gas has been passed through a free radical source 15A for turning the hydrogen gas into free radicals, whereby a mixed gas of the hydrogen gas and the hydrogen radicals has been introduced into the chamber. FIG. 5 shows a relationship between the flow rate of the hydrogen gas and the etching rate of Sn on the surface of the EUV collector mirror 12. The vertical axis represents the etching rate (nm/min) of Sn, and the horizontal axis represents the mass flow rate (sccm) of the hydrogen gas. Note that the etching rate of Sn was measured under the condition where the pressure inside the chamber was at or below 10 Pa. When, for example, the deposition rate of Sn on the surface of the EUV collector mirror 12 is at 0.04 nm/min, adjusting the mass flow rate of the hydrogen gas to a range between 25 sccm and 131 sccm prevents Sn from being deposited on the surface of the EUV collector mirror 12. Next, a description will be provided below that as the pressure inside the chamber 10 increases, the trapping effect of ions by the magnetic field decreases. When the gas pressure inside the chamber 10 (hereinafter referred to as chamber pressure) is increased, collision between the ionized debris traveling along the magnetic force lines MF and molecules (or atoms) inside the chamber 10 occurs frequently. As a result, the ionized debris that is trapped in the magnetic field and travels therein deviates largely from the path extending along the magnetic force lines MF, and diffuses to the outside with respect to the central line of the magnetic field. That is, as the chamber pressure is increased, the cross sectional area of the ion flow FL flowing in the direction of the magnetic field increases. As the chamber pressure is further increased, the stopping distance of the traveling ions becomes shorter, whereby the ions cannot be collected. Here, the diffusion width of the ionized debris caused by colliding with the molecules (or atoms) depends on the atmosphere gas inside the chamber 10, particularly on the type and the pressure of the etching gas introduced into the chamber 10. FIG. 6 shows a result of a Monte Carlo simulation (SRIM) for finding a trajectory of an Sn ion when the Sn ion of 1 key enters the hydrogen gas atmosphere where the gas pressure is 1 Pa and the magnetic field having the magnetic flux density of 0.6 T in the direction of entry. Further, FIG. 7 shows a result of a Monte Carlo simulation (SRIM) for finding a trajectory of an Sn ion when the Sn ion of 0.01 keV enters the Xe gas atmosphere where the gas pressure is 1 Pa and the magnetic field having the magnetic flux density of 0.6 T in the direction of entry. Note that in FIG. 6 and FIG. 7, the horizontal axis represents the distance in the direction of entry, that is, the moving distance in the direction of entry, and the vertical axis represents the moving distance in the direction perpendicular to the direction of entry, that is, the diffusion width σ. First, as shown in FIG. 6, the Sn ion that has entered the hydrogen gas atmosphere is not diffused substantially when the moving distance is around a few hundred millimeters, but as it travels in the hydrogen gas, the diffusion width σ increases. On the other hand, the Sn ion that has entered the Xe gas atmosphere is largely diffused immediately after it has entered the Xe gas, and it stops without moving much. Further, while the stopping distance of the Sn ion in the direction of entry in the hydrogen gas is approximately 6 m, the stopping distance of the Sn ion in the direction of entry in the Xe gas is as short as approximately 300 mm at most. This is because the hydrogen molecule has sufficiently small mass with respect to the Sn ion, and thus the energy loss at collision is small; whereas, the Xe atom has substantially the same mass as the Sn ion, and thus the energy loss at collision is large. Based on these, it is contemplated that in order to prevent the ionized debris from being diffused, a gas that is relatively light in mass, such as hydrogen gas, is preferably used as the etching gas. Further, FIG. 8 shows a relationship between the chamber pressure and the diffusion width of the Sn ion found by a Monte Carlo simulation (SRIM) in accordance with the first embodiment. In FIG. 8, a line Lh1 represents the dependency of the diffusion width σ on the chamber pressure in a case where hydrogen gas is used as the atmosphere gas and the moving distance of the Sn ion in the direction of entry is 100 mm, and a line Lh2 represents the dependency of the diffusion width σ on the chamber pressure in a case where hydrogen gas is used as the atmosphere gas and the moving distance of the Sn ion in the direction of entry is 200 mm. Further, a line Lxe1 represents the dependency of the diffusion width σ on the chamber pressure in a case where Xe gas is used as the atmosphere gas and the moving distance of the Sn ion in the direction of entry is 100 mm, and a line Lxe2 represents the dependency of the diffusion width σ on the chamber pressure in a case where Xe gas is used as the atmosphere gas and the moving distance of the Sn ion in the direction of entry is 200 mm. As shown in FIG. 8, when the hydrogen gas is used as the atmosphere gas, the diffusion width σ is at or below 1 mm when the chamber pressure is approximately at or below 6×10−1 Pa in the case where the moving distance in the direction of entry is 100 mm, or when the chamber pressure is approximately at or below 3 Pa in the case where the moving distance is 200 mm. On the other hand, when the Xe gas is used as the atmosphere gas, the diffusion width σ is at or below 1 mm when the chamber pressure is approximately at or below 3.5×10−4 Pa in the case where the moving distance in the direction of entry is 100 mm, or when the chamber pressure is approximately at or below 8×10−5 Pa when the moving distance is 200 mm. Here, a mass ratio of hydrogen gas to an Sn ion is sufficiently small with respect to a mass ratio of Xe gas to an Sn ion. Accordingly, using hydrogen gas as the atmosphere gas makes it possible to obtain substantially the same diffusion width σ as the case where the Xe gas is used as the atmosphere gas even when the chamber pressure is increased approximately by four orders. That is, using hydrogen gas as the atmosphere gas makes it possible to increase the chamber pressure without increasing the diffusion width σ, in comparison to the case where the Xe gas is used. Based on the above, it is contemplated that, in order to remove the deposited Sn more efficiently, hydrogen gas or hydrogen radical is preferably used as the etching gas. Note that Xe gas does not react with Sn; thus, Xe gas cannot be used as the etching gas. However, Xe gas is effective in stopping Sn debris having high energy. Thus, in the first embodiment, the magnetic flux density B of the magnetic field generated around the plasma generation site P1 and the positional relationship of the EUV collector mirror 12 and the ion collection unit 13 with respect to the plasma generation site P1 are arranged as follows. FIG. 9 shows an exemplary positional relationship among the plasma generation site P1, the EUV collector mirror 12, and the ion collection unit 13 in accordance with the first embodiment. In the example shown in FIG. 9, hydrogen gas is used as the etching gas, Sn is used as the target material, the magnetic flux density B of the magnetic field generated around the plasma generation site P1 is 0.6 T, a radius r1 of the EUV collector mirror 12 is 200 mm, a distance l1 between an edge of the ion flow FL in the direction perpendicular to the direction of the magnetic field and an edge of the EUV collector mirror 12 is 6 mm. Note that when the magnetic flux density is 0.6 T, a radius r2 of the ion flow FL in a vacuum state is 25 mm. Further, the diffusion width of the Sn ion including a safety factor is designated by W1h. Under this condition, when, for example, the safety factor is 3 (W1h=3σ) and the chamber pressure is at 10 Pa, the probability of an Sn ion being incident on the EUV collector mirror 12 is as small as 1.35×10−3. Further, when, for example, the safety factor is 6 (W1h=6σ) and the chamber pressure is 4 Pa, the probability of an Sn ion being incident on the EUV collector mirror 12 is 1.78×10−10, which is thus further improved. Furthermore, when, for example, the safety factor is 9 (W1h=9σ) and the chamber pressure is 2 Pa, the probability of an Sn ion being incident on the EUV collector mirror 12 is 1.13×10−19, which is negligibly small. Under the condition shown in FIG. 9, ionized debris incident on the ion collection unit 13 and ionized debris incident on the EUV collector mirror 12 were measured. FIG. 10 is a diagram for describing an experiment in which an amount of ionized debris incident on the ion collection unit 13 and on the EUV collector mirror 12 shown in FIG. 9 is measured. As shown in FIG. 10, in this experiment, a Faraday cup F1 is disposed in place of the ion collection unit 13, and a Faraday cup F2 is disposed in place of the edge of the EUV collector mirror 12. The Faraday cup F1 may be disposed at a location corresponding to the edge of the EUV collector mirror 12, as shown in FIG. 10. Further, FIG. 11 shows a temporal change in current flowing into the Faraday cup (F1) disposed on an axis of the magnetic field in a case where the interior of the chamber is kept in vacuum and in a case where the interior of the chamber is under low-pressure hydrogen gas atmosphere as shown in FIG. 10. FIG. 12 shows a temporal change in current flowing into the Faraday cup (F2) disposed off an axis of the magnetic field in a case where the interior of the chamber is kept in vacuum and in a case where the interior of the chamber is under low-pressure hydrogen gas atmosphere as shown in FIG. 10. A line L11 shown in FIG. 11 represents a detected waveform in the Faraday cup F1 in a case where the chamber is kept in vacuum and the magnetic field with the magnetic flux density of 0.6 T is generated. Under this condition, relatively large, short pulsed current was detected in the Faraday cup F1. On the other hand, a line L22 shown in FIG. 12 represents a detected waveform in the Faraday cup F2 under the same condition as the above. Under this condition, pulsed current was not detected in the Faraday cup F2. Based on these, keeping the chamber in vacuum allows almost all ions to reach the Faraday cup F1. On the other hand, the ions do not reach the Faraday cup F2 corresponding to the EUV collector mirror 12. This indicates that the ionized Sn debris having high energy is not incident on the EUV collector mirror 12. In other words, compared to the current detected in the Faraday cup F1, the current detected in the Faraday cup F2 is sufficiently small, which indicates that almost all ionized debris can be collected in the Faraday cup F1 (corresponding to ion collection unit 13). Meanwhile, a line L12 shown in FIG. 11 represents a detected waveform in the Faraday cup F1 in a case where hydrogen gas is introduced into the chamber 10 to increase the pressure thereinside and the magnetic field having the magnetic flux density of 0.6 T is generated. Under this condition, relatively large, short pulsed current was detected in the Faraday cup F1. On the other hand, a line L21 shown in FIG. 12 represents a detected waveform in the Faraday cup F2 under the same condition as the above. Under this condition, small, pulsed current was detected in the Faraday cup F2. This case corresponds to a case where the ionized Sn debris is incident on the EUV collector mirror. FIG. 13 shows values of integral of current detected in each Faraday cup in a case where hydrogen gas pressure inside the chamber is varied as shown in FIG. 10. As shown in FIG. 13, when the hydrogen gas pressure is below 8 Pa, the values of integral of the current detected in the Faraday cups F1 and F2 shown in FIG. 10 do not change substantially. This indicates that the positional relationship with respect to the magnetic flux density B as shown in FIG. 9 enables to collect almost all generated ionized debris in the ion collection unit 13 while the chamber pressure containing hydrogen gas does not exceed at least 8 Pa. Accordingly, when hydrogen gas is used as the etching gas, keeping the hydrogen gas pressure inside the chamber 10 at least at or below 8 Pa allows to collect the ionized debris generated around the plasma generation site P1 using the magnetic field and to prevent a layer of the target material from being formed on the surface of the optical element inside the chamber 10. The above experimental result matched with the simulation result. Next, FIG. 14 is a graph showing a relationship between an averaged stopping distance of Sn ions and the hydrogen gas pressure. The vertical axis represents the averaged stopping distance (mm), and the horizontal axis represents the hydrogen gas pressure (Pa). The higher the hydrogen gas pressure is, the shorter the averaged stopping distance d1 is. For example, in order to trap Sn ions in the magnetic field at a position 200 mm away from the plasma generation site P1 to collect Sn, the hydrogen gas pressure needs to be at or below 15 Pa. Further, in order to trap the Sn ions in the magnetic field at a position 46 mm away from the plasma generation site P1 to collect Sn, the hydrogen gas pressure needs to be at or below 100 Pa. As described above, in the first embodiment, the distance between the optical element and the edge of the ion flow FL is adjusted such that the optical element such as the EUV collector mirror 12 is not in a range where the ionized debris is diffused, and the pressure of the etching gas inside the chamber 10 is optimized. With this, the ionized debris generated around the plasma generation site P1 can be collected using the magnetic field while a layer of the target material can be prevented from being formed on the surface of the optical element inside the chamber 10. In the above description, a case where the diffused ionized debris is prevented from being incident on the EUV collector mirror 12 has been exemplified. However, the present disclosure is not limited thereto. FIG. 15 illustrates an exemplary positional relationship among the plasma generation site, the EUV collector mirror, and the nozzle of the droplet generator in accordance with the first embodiment. As shown in FIG. 15, for example, the tip of the nozzle 16a of the droplet generator 16 is preferably positioned as close to the plasma generation site P1 as possible so that the droplet generator 16 can supply the droplet D with high precision and at favorable timing to the plasma generation site P1. However, if the ionized debris is deposited onto the tip of the nozzle 16a, the droplet D may not be outputted with high precision through the tip of the nozzle 16a. Thus, in the first embodiment, as shown in FIG. 15, a distance 12 between the edge of the ion flow FL and the tip of the nozzle 16a is regulated such that the tip of the nozzle 16a is not disposed in a range where the ionized debris is diffused, and the pressure of the etching gas inside the chamber 10 is optimized. Note that in this example, the direction into which the target is outputted is inclined with respect to the direction of the magnetic field. Without being limited to this example, however, the arrangement may be such that the direction into which the target is outputted is substantially perpendicular to the direction of the magnetic field, and the ionized debris does not reach the target material outlet port of the nozzle 16a. Further, FIG. 16 illustrates an exemplary positional relationship among the plasma generation site, the EUV collector mirror, and a measuring unit in accordance with the first embodiment. A measuring unit M1 for measuring light intensity or the like of the EUV light L2 emitted at the plasma generation site P1 is also preferably disposed as close to the plasma generation site P1 as possible in order to improve its measuring accuracy. Thus, in the first embodiment, a distance 13 between the edge of the ion flow FL and the tip of the measuring unit M1 is regulated, as shown in FIG. 16, such that the tip of the measuring unit M1 is not disposed in a range where the ionized debris is diffused, and the pressure of the etching gas inside the chamber 10 is optimized. Note that in this embodiment, a case of the measuring unit M1 for measuring the light intensity was exemplified, but without being limited to thereto, the embodiment can be applied to any sensor disposed inside the chamber. It may be applied, for example, to a detector for detecting the position of the target. Furthermore, FIG. 17 illustrates an exemplary positional relationship among the plasma generation site, the EUV collector mirror, and an etching gas introduction unit in accordance with the first embodiment. The ionized debris generated around the plasma generation site P1 tends to be deposited onto the optical element disposed around the plasma generation site P1. Thus, the etching gas introduction unit 15 for introducing the etching gas into the chamber 10 is preferably disposed around the plasma generation site P1. Therefore, in the first embodiment, a distance 14 between the edge of the ion flow FL and the tip of the etching gas introduction unit 15 is regulated, as shown in FIG. 17, such that the tip of the etching gas introduction unit 15 is not disposed in a range where the ionized debris is diffused, and the pressure of the etching gas inside the chamber 10 is optimized. Note that the etching gas introduction unit 15 may be a radical introduction port for introducing the hydrogen radial or the like into the chamber 10. FIG. 18 illustrates an exemplary positional relationship among the plasma generation site, the EUV collector mirror, and a free radical source in accordance with the first embodiment. As described above, the etching gas introduced into the chamber 10 is preferably a free radical that is highly reactive with the target material. Thus, in the first embodiment, as shown in FIG. 18, the free radical source 15A may be disposed at the tip of the etching gas introduction unit 15. With this, the etching gas supplied from the etching gas introduction unit 15 can be turned into a free radical by the free radical source 15A and the etching gas that has been turned into a free radical can be supplied into the chamber 10. However, as described with reference to FIG. 17, a part (free radical source 15A) for introducing the etching gas into the chamber 10 is preferably disposed around the plasma generation site P1. Thus, in the first embodiment, a distance 15 between the edge of the ion flow FL and the tip of the free radical source 15A is regulated, as shown in FIG. 18, such that the tip of the free radical source 15A is not disposed in a range where the ionized debris is diffused, and the pressure of the etching gas inside the chamber 10 is optimized. FIG. 19 illustrates an exemplary positional relationship among the plasma generation site, the EUV collector mirror, and an electrostatic suction type droplet generator in accordance with the first embodiment. The droplet generator 16 in accordance with the first embodiment may be replaced by an electrostatic suction type droplet generator 16A shown in FIG. 20. FIG. 20 illustrates a schematic configuration of the electrostatic suction type droplet generator shown in FIG. 19. As shown in FIG. 20, the electrostatic suction type droplet generator 16A includes a tank 16-1 for storing liquid state molten Sn thereinside, an electrode 16c for grounding the molten Sn inside the nozzle 16a projecting from the tank 16-1, and a suction electrode 16b disposed to face the electrode 16c with a space therebetween. The molten Sn inside the tank 16-1 may be heated to or above its melting point by a heater 16-2 provided around the tank 16-1, for example. Further, the suction electrode 16b is spaced from the electrode 16c by an insulator 16d and is fixed thereto. With this configuration, applying voltage of a several kV to the suction electrode 16b in pulses allows to pull out the droplet D through the tip of the nozzle 16a on demand. Note that even in a case where the electrostatic suction type droplet generator 16A is used, similarly to the case where the droplet generator 16 is used, the tip of the nozzle 16a is preferably disposed as close to the plasma generation site P1 as possible. Thus, in the first embodiment, a distance 16 between the edge of the ion flow FL and the tip of the suction electrode 16b is regulated, as shown in FIG. 19, such that the suction electrode 16b positioned to face the nozzle 16a is not disposed in a range where the ionized debris is diffused, and the pressure of the etching gas inside the chamber 10 is optimized. Further, FIG. 21 illustrates an exemplary positional relationship among the plasma generation site, the EUV collector mirror, and an electrostatic suction and acceleration type droplet generator in accordance with the first embodiment. The droplet generator 16 in accordance with the first embodiment may be replaced by an electrostatic suction and acceleration type droplet generator 16B shown in FIG. 22. FIG. 22 illustrates a schematic configuration of the electrostatic suction and acceleration type droplet generator shown in FIG. 21. As shown in FIG. 22, the electrostatic suction and acceleration type droplet generator 16B includes, in addition to the configuration similar to that of the electrostatic suction type droplet generator 16A shown in FIG. 20, an acceleration electrode 16e disposed to face the suction electrode 16b. With this configuration, applying voltage of a several kV to the suction electrode 16b in pulses and applying voltage of a several ten kV or above to the acceleration electrode 16e enables to accelerate the droplet D pulled out through the tip of the nozzle 16a on demand. In a case where the electrostatic suction type droplet generator 16A is used as described above, or in a case where the droplet is charged and is accelerated, high voltage of a several kV needs to be applied to an electrode facing a nozzle opening of the droplet generator. However, when the gas pressure inside the chamber 10 is high, insulation breakdown occurs in the gas; thus, high voltage cannot be applied. For this reason, when there is an element to which high voltage is applied, the maximum value of the gas pressure needs to be limited. For example, when the above-described electrostatic suction type droplet generator 16A or the acceleration electrode 16e is used, the chamber pressure is preferably kept approximately at or below 0.2 Pa. Note that even in a case where the electrostatic suction and acceleration type droplet generator 163 is used, similarly to the case where the droplet generator 16 is used, the tip of the nozzle 16a is preferably disposed as close to the plasma generation site P1 as possible. Thus, in the first embodiment, a distance 17 between the edge of the ion flow FL and the acceleration electrode 16e is regulated, as shown in FIG. 21, such that the acceleration electrode 16e positioned to away from the nozzle 16a is not disposed in a range where the ionized debris is diffused, and the pressure of the etching gas inside the chamber 10 is optimized. As described above, the optical element in accordance with the first embodiment includes, without being limited to the EUV collector mirror 12, various optical elements disposed inside the chamber 10. Further, elements such as the target generation nozzle through which the target is generated, various sensors, and the etching gas introduction unit or the like disposed inside the chamber are also included. Next, an EUV light generation apparatus in accordance with a second embodiment of the present disclosure will be described in detail with reference to the drawing. FIG. 23 is a sectional view illustrating a schematic configuration of the EUV light generation apparatus in accordance with the second embodiment. As shown in FIG. 23, an EUV light generation apparatus 2 in accordance with the second embodiment includes, in addition to the configuration similar to that of the EUV light generation apparatus 1 in accordance with the first embodiment, an ion sensor 21 disposed close to the ion collection unit 13, a mass flow controller (MFC) 22 for controlling the flow rate of the etching gas introduced into the chamber 10 from the etching gas introduction unit 15, the MFC 22 being disposed between the gas tank 24 for storing the etching gas and the etching gas introduction unit 15, a controller 23 for controlling the MFC 22 based on the amount of ions detected at the ion sensor 21, an exhaust pump 20 for discharging the gas inside the chamber 10, and a pressure sensor 25 for detecting the chamber pressure. Note that the gas tank 24 was omitted in the above-described first embodiment, but it is also provided to the EUV light generation apparatus 1. Further, the MFC and the controller 23 regulates the flow rate of the etching gas such that the gas pressure inside the chamber 10 is at the gas pressure at which the diffusion width of the ionized debris is shorter than the distance between the edge of the ion flow FL in which the ionized debris travels while being trapped in the magnetic field and the optical element, and controls the exhaust pump 20 to control the discharge rate of the gas to be discharged. The ion sensor 21 detects the amount of the ionized debris that did not enter the ion collection unit 13 because it was diffused by the gas (mainly etching gas) inside the chamber 10. Thus, the ion sensor 21 is disposed next to the ion collection unit 13 and as close to the ion collection unit 13 as possible. The controller 23 controls the MFC 22 when an ion is detected at the ion sensor 21 or when the amount of ions detected at the ion sensor 21 exceeds a predetermined threshold value, whereby the flow rate of the etching gas flowing into the etching gas introduction unit 15 from the gas tank 24 is temporarily reduced. With this, the amount of the etching gas introduced into the chamber 10 can be adjusted to lower the chamber pressure. As a result, the diffusion width of the ionized debris trapped in the magnetic field is reduced, and the amount of ionized debris that does not flow into the ion collection unit 13 is reduced. Alternatively, the flow rate at the mass flow controller (MFC) 22 for controlling the flow rate of the etching gas is kept constant, and the discharge rate at the exhaust pump 20 may be controlled. This way, by optimizing the gas pressure inside the chamber 10 based on the actual amount of the ionized debris that does not flow into the ion collection unit 13, the ionized debris generated around the plasma generation site P1 can be collected into the ion collection unit 13 more reliably. With this, the ionized debris generated around the plasma generation site P1 can be collected using the magnetic field, and a layer of the target material can be prevented from being formed on the surface of the optical element inside the chamber 10 more reliably. Note that in the above description, the ion sensor 21 is disposed close to the ion collection unit 13, and the amount of the ionized debris that does not flow into the ion collection unit 13 is detected. However, without being limited thereto, the ion sensor 21 connected to the controller 23 may be disposed close to various optical elements, such as the EUV collector mirror 12, the nozzle 16a of the droplet generator 16, the measuring unit M1, the etching gas introduction unit 15, or the free radical source 15A, whereby the amount of the ionized debris incident on these optical elements may be detected. Based on the detected result, the chamber pressure may be controlled using at least either one of the MFC 22 and the exhaust pump 20. Next, an EUV light generation apparatus in accordance with a third embodiment will be described in detail with reference to the drawing. FIG. 24 is a sectional view illustrating a schematic configuration of the EUV light generation apparatus in accordance with the third embodiment. As shown in FIG. 24, an EUV light generation apparatus 3 in accordance with the third embodiment includes, in addition to the configuration similar to that of the EUV light generation apparatus 2 in accordance with the above-described second embodiment, a magnet controller 31 connected to each of the pair of magnets 14. The magnet controller 31 and the controller 23 function as a magnetic field intensity control unit for controlling the intensity of the magnetic field such that the magnetic flux density of the magnetic field generated around the plasma generation site P1 is the magnetic flux density at which the diffusion width of the ionized debris is shorter than the distance between the edge of the ion flow FL flowing while being trapped in the magnetic field and the optical element. In this case, a correlation between current applied to the coil 14a and the magnetic flux density of the magnetic field generated around the plasma generation site P1 may be calculated in advance, and current applied to the coil 14a may be controlled so that the magnetic field of a desired magnetic flux density can be generated. Note that a magnetic sensor (not shown) may be provided around the magnet 14 to measure the magnetic flux density. When the magnetic sensor is provided, the magnet controller 31 may control the intensity of the magnetic field based on the magnetic flux density value detected by the magnetic sensor. In this case, a correlation between a magnetic flux density at a position where the magnetic flux density is measured by the magnet sensor and a magnetic flux density of the magnetic field generated around the plasma generation site P1 may be calculated in advance, and current applied to the coil 14a may be controlled so that the magnetic field of a desired magnetic flux density can be generated. Under the control of the controller 23, the magnet controller 31 controls the intensity of the magnetic field generated by the pair of magnets 14. That is, the controller 23 controls the MFC 22 to temporarily reduce the flow rate of the etching gas flowing into the etching gas introduction unit 15 from the gas tank 24 and controls the magnet controller 31 when an ion is detected at the ion sensor 21 or when the amount of ions detected at the ion sensor 21 exceeds a predetermined threshold value, whereby the intensity of the magnetic field generated around the plasma generation site P1 is increased. With this, the amount of the etching gas introduced into the chamber 10 is adjusted to lower the chamber pressure and to increase the magnetic flux density, whereby the diffusion width of the ionized debris can be reduced. As a result, the diffusion width of the ionized debris trapped in the magnetic field is reduced, and the amount of the ionized debris that does not flow into the ion collection unit 13 is reduced. This way, by optimizing the gas pressure inside the chamber 10 and the magnetic flux density based on the actual amount of the ionized debris that does not flow into the ion collection unit 13, the ionized debris generated around the plasma generation site P1 can be collected into the ion collection unit 13 more reliably. With this, the ionized debris generated around the plasma generation site P1 can be collected using the magnetic field, and a layer of the target material can be prevented from being formed on the surface of the optical element inside the chamber 10 more reliably. Note that in the above description, the ion sensor 21 is disposed close to the ion collection unit 13 to detect the amount of the ionized debris that does not flow into the ion collection unit 13. However, without being limited thereto, the ion sensor 21 connected to the controller 23 may be disposed close to various optical elements, such as the EUV collector mirror 12, the nozzle 16a of the droplet generator 16, the measuring unit M1, the etching gas introduction unit 15, or the free radical source 15A, whereby the amount of the ionized debris incident on these optical elements may be detected. Next, an EUV light generation apparatus in accordance with a fourth embodiment will be described in detail with reference to the drawings. FIGS. 25 and 26 are sectional views each illustrating a schematic configuration of the EUV light generation apparatus in accordance with the fourth embodiment. FIG. 26 is a sectional view taken along a plane containing the axis AX and along a different plane from that shown in FIG. 25. As shown in FIGS. 25 and 26, an EUV light generation apparatus 4 in accordance with the fourth embodiment includes a configuration similar to that of any one of the EUV light generation apparatuses of the above first through third embodiments, but the coils 14a for generating the magnetic field around the plasma generation site P1 are disposed inside the chamber 10. Note that the chamber 10 is connected to the exposure device connection unit 19 via a gate valve 49 for maintaining the airtightness of the chamber 10. Further, the EUV light generation apparatus 4 includes a magnetic core 42 extending cylindrically toward the plasma generation site P1 from a bore of each coil 14a. This makes it possible to generate a strong magnetic field around the plasma generation site P1. As a result, a strong magnetic field around the plasma generation site P1 can be generated while effects of the magnetic field on the measuring devices or the like therearound are minimized. Further, the central line of magnetic force of the magnetic field generated by the magnetic cores 42 of which tips face each other with the plasma generation site P1 therebetween passes through the plasma generation site P1, and the direction thereof coincides with axes of the cylindrical magnetic cores 42. Accordingly, the ion collection unit 13 is disposed at the center of the bore of each coil 14a and inside each magnetic core 42. Further, a diameter of an opening of each magnetic core 42 toward the plasma generation site P1 is at least larger than the value in which a diffusion width of the ion debris is added to the diameter of the ion flow FL, similarly to the positional relationship between the edge of the ion flow FL and the edge of the EUV collector mirror 12. With this, the ionized debris generated around the plasma generation site P1 can reliably be introduced into the interior of the magnetic core 42. As a result, the ionized debris generated around the plasma generation site P1 can be trapped in the magnetic field generated intensively around the plasma generation site P1 and thereafter be collected into the ion collection unit 13 disposed at the bottom of the magnetic core 42. With the above configuration, the ionized debris can be trapped using a stronger magnetic field; thus, even when the etching gas pressure inside the chamber 10 is increased, the diffusion width of the ionized debris can be minimized. As a result, in a configuration where the ionized debris generated around the plasma generation site P1 is collected using the magnetic field, a layer of the target material can more reliably be prevented from being formed on the surface of the optical element inside the chamber 10. In this embodiment, the magnetic cores 42 can be disposed close to the plasma generation site P1; thus, even when the etching gas pressure is increased, the ionized debris can be trapped. For example, in a case where an etching gas including hydrogen gas or a hydrogen radical is used, even when the chamber pressure is increased up to 100 Pa, the ionized debris can be trapped and collected, and Sn deposited on the EUV collector mirror 12 can etched. Further, the magnetic cores 42 disposed inside the chamber 10 extend into an obscuration region E1 from the coils 14a disposed outside the obscuration region E1. Here, FIG. 27 illustrates a far field pattern formed on A-A plane by the EUV light shown in FIG. 26. As shown in FIG. 27, the EUV light L2 includes the obscuration region E1 along a plane perpendicular to the axis AX. The obscuration region E1 refers to a region corresponding to an angular range where the EUV light L2 collected by the EUV collector mirror 12 is not used in the exposure device 100. In the following description, a three-dimensional volumetric region contained in the angular range of the EUV light L2 which is not used in the exposure device 100 is referred to as the obscuration region E1. As described above, the EUV light L2 in the obscuration region E1 is not used for exposure in the exposure device 100. Thus, even when the EUV light L2 in the obscuration region E1 is not inputted into the exposure device 100, exposure performance or throughput of the exposure device 100 is not affected at all. Therefore, in the fourth embodiment, as shown in FIG. 28, the magnetic cores 42 are made to extend into the obscuration region E1. With this, the tips of the magnetic cores 42 can be disposed even closer to the plasma generation site P1. As a result, a stronger magnetic field can be generated in a smaller region around the plasma generation site P1. Note that FIG. 28 illustrates an exemplary positional relationship between the magnetic cores and the obscuration region in accordance with the fourth embodiment. Further, as shown in FIG. 29, the magnetic core 42 that extends into the obscuration region E1, for example, is preferably coated with a coating film 42c made of a material that is less likely to be sputtered by the ionized debris, the material including tungsten (W), tin (Sn), ruthenium (Ru), molybdenum (Mo), silicon (Si), carbon (C) or the like. Further, it may be coated with Cu or Ti that is highly wettable with Sn. With this, the magnetic core 42 is prevented from being sputtered, whereby debris of the material of the magnetic core 42 is prevented from being generated inside the chamber 10. Note that FIG. 29 illustrates an exemplary positional relationship among the ion flow, the magnetic core, the coil, and the ion collection unit in accordance with the fourth embodiment. Further, as shown in FIG. 29, the ion collection unit 13 is also preferably coated with a coating film 13c made of a material similar to that for the coating film 42c. With this, the ion collection unit 13 can be prevented from being sputtered. Other configurations are the similar to those of any one of the above-described first through third embodiments; thus, the duplicate description thereof is omitted here. Note that a driver laser LD10 shown in FIGS. 25 and 26 is a configuration of which description is omitted in each of the above embodiments. However, it can be provided in each embodiment, and it serves as a light source of the laser beam L1. An EUV light generation apparatus in accordance with a fifth embodiment of the present disclosure will be described in detail with reference to the drawing. FIG. 30 is a sectional view illustrating a schematic configuration of the EUV light generation apparatus in accordance with the fifth embodiment. As shown in FIG. 30, an EUV light source device 5 in accordance with the fifth embodiment includes a similar configuration as that of the EUV light generation apparatus 2 in accordance with the above-described second embodiment, but includes only one magnet 14. That is, in the fifth embodiment, a case where the magnetic field is generated around the plasma generation site P1 by a single coil configured of the one magnet 14 will be exemplified. When plasma is generated around the plasma generation site P1, ions are trapped in the magnetic field and travel in the direction of the magnetic field. The magnetic flux density at the side of the magnet 14 is higher with respect to the plasma generation site P1. On the other hand, the magnetic flux density at the side where the magnet 14 is not disposed is lower with respect to the plasma generation site P1. Thus, in accordance with the state of the magnetic flux density, the ion collection unit at the side where the magnet 14 is not disposed is replaced by an ion collection unit 13A having a larger opening than the ion collection unit 13 at the side where the magnet 14. In the fifth embodiment configured as such, the ion collection unit 13 and the ion collection unit 13A are disposed with their opening size being adjusted. In the fifth embodiment, however, is not limited to the case where the ion collection unit 13 and the ion collection unit 13A have differing opening diameters, but the ion collection unit 13 may be disposed in place of the ion collection unit 13A. In such configuration, the controller 23 detects the pressure inside the chamber 10 by a signal from the pressure sensor 25 to thereby control the flow rate of the etching gas with the MFC 22, or controls the discharge rate of the exhaust pump 20 to thereby control the etching gas pressure to desired pressure. The above-described embodiments and the modifications thereof are merely examples for implementing the present disclosure, and the present disclosure is not limited thereto. Further, making various modifications in accordance with the specification is within the scope of the present disclosure, and it is apparent that the various other embodiments can be made from the above description without departing from the scope of the present disclosure. For example, it is needless to state that the modifications indicated for each of the embodiments can be applied to the other embodiments. |
|
summary | ||
059490837 | summary | FIELD OF THE INVENTION The invention concerns a container for transporting or storing nuclear fuel assemblies said container comprising a forged metal body, usually of forged steel, which delimits a cavity into which the assemblies are placed. The invention also concerns its method of manufacture. DESCRIPTION OF THE RELATED ART Nuclear fuel assemblies are generally prismatic or cylindrical, usually with a square cross-section (for example, PWR, BWR, . . . ), occasionally with a hexagonal cross-section (VVER, . . .). and occasionally with a circular section (CANDU, RBMK, . . . ). One problem with which the designer of the container is faced is that of being able to house the maximum number of assemblies in the container cavity while complying with current regulations; in particular, the container must have: shielding which is sufficient against the radiation emitted by the radioactive material contents; PA1 mechanical strength and a seal which are sufficient to ensure containment of the radioactive material contents, even in the event of an accident; PA1 sufficient thermal conductivity to conduct the heat which may be released by the radioactive assemblies housed in the cavity towards the body of the container, in order to limit the temperature reached. In addition, the weight of the container and its overall dimensions must be limited so as to be compatible with the equipment in the operational installations (reactors, intermediate storage installations, reprocessing plants, transportation means). A container with a cavity whose cross-sectional shape is adapted to that of the fuel assemblies and whose body shape is such that side by side storage of a plurality of containers takes up a minimum of space, is of particular advantage. From this viewpoint, it is not entirely satisfactory to use a container having, for example, a cavity with a circular cross-section to house fuel assemblies of square cross-section. Similarly, containers whose outer body surface is of circular cross-section are not always satisfactory. Metal containers can be constructed using a number of technologies: cast iron containers (casting), multi-layered steel/steel containers (rolled-welded), lead containers (lead cast between two rolled-welded steel enclosures), forged steel containers (shell forged and then lathe turned). These technologies which are based on procedures involving rotation about an axis (rolling of sheet metal, forging of shells, lathe turning, etc. . . ) are not suitable for putting the above idea of compact storage into practice. Technologies based on processes such as casting can in principle produce containers with non-circular cross-sections but for reasons connected with the constitution of cast material it is often preferable to use circular containers which allow better control over cooling conditions. For example, British patent GB 2 003 783 describes a cast iron or steel container for transporting and storing radioactive waste having a non-circular cross-section. It will also be noted that French patent FR 2 563 652 describes a shell comprising two steel walls with a maximum thickness of 4 mm, between which a thin neutron-absorbing screen is disposed; the square shape of the shell is produced by plastic deformation of the steel walls. We have thus sought to develop containers with improved weight and overall dimensions relative to the number of fuel assemblies it contains, while observing shielding, thermal conductivity and mechanical strength requirements. SUMMARY OF THE INVENTION The invention concerns a container for nuclear fuel assemblies, comprising a thick cylindrical body of forged steel which delimits a cavity for housing said nuclear fuel assemblies, said cavity being able to be hermetically sealed at its two ends by plugs which are also formed of metal, characterized in that the cross-section of the cylindrical body is non-circular. The container according to the invention thus has a cylindrical metal body with a non-circular cross section. In other words, the cross-section has the appearance of a ring whose inner and outer perimeters are not circular but generally contain straight segments; the perimeters can, for example, be in the shape of squares or other regular concentric polygons whose corners can be rounded. It is formed by taking a thick metal shell with a circular cross section which delimits an inner cavity in which the assemblies are housed, and forming one or more flat surfaces on the outer and/or inner wall of the shell, the flat surfaces usually being symmetrically disposed on the outer and inner perimeters, and facing one another. For this reason these flat surfaces are formed by grinding so as to extend over the entire height of the outer wall, and/or at least one crescent-shaped section is inserted into the cavity which matches the shape of the inner wall and is fixed thereto. These crescent-shaped sections have a cross-section which comprises an arc of a circle with the same diameter as the inner wall of the cavity and a chord which subtends the arc of the circle and which thus corresponds to a flat portion of the inner wall of the cavity. There can be 2, 4 or 6 flat surfaces; the outer or inner perimeter of the ring has a square or rectangular shape when there are 4 flat surfaces, or a hexagonal shape when there are 6 flat surfaces. The thickness of the cylindrical metal body with a noncircular cross-section according to the invention completely satisfies shielding standards. It is usually several tens of centimeters thick. It can be seen that the shape of the inner cavity can be adapted to the type of fuel assemblies which it is to house. Thus, when the assemblies have a square cross-section, for example, a cavity is preferably selected which has a square or rectangular cross-section, usually with rounded corners; this makes it possible to increase its filling coefficient (less lost space than in a cavity with a circular cross-section). By forming flat surfaces on the outer wall of the cylindrical body, which flat surfaces usually face the flat portions of the inner wall, the weight and overall dimensions of the container are simultaneously reduced while adequate shielding and mechanical strength is maintained; due to their outer shape the storage density per m.sup.2 of the containers is increased. The cavity inside the cylindrical metal body is usually sealed at its two ends, one which is sealed by a fixed base, attached, for example, by welding with or without hooping, the other by means of a removable cover. The initial shell with a circular cross-section which must be modified to produce the cylindrical body according to the invention is generally forged steel-based. Consequently, the cylindrical body of the container according to the invention is of the same composition. Thus, the outer and inner walls of the forged steel shell are lathe turned coaxially, initially to produce a cylindrical body of circular cross section, then the outer wall is ground to obtain at least one flat surface extending over the entire height of the shell, and preferably 2, 4 or 6 parallel paired flat surfaces which are also symmetrically paired with respect to the axis of the cylindrical body. There are preferably 2, 4 or 6 crescent-shaped sections, which may or may not be identical, and their chords are in parallel pairs and also symmetrically paired with respect to the axis of the cylindrical body. The inner wall of the shell and the crescents can be covered with a metallic coating, for example a Al-Zn coating produced using Schoop's metal spraying process. |
045499853 | description | DETAILED DESCRIPTION OF THE INVENTION This invention comprises a method for processing an acid solution containing dissolved metal compounds such as spent acid electrolyte from an electrolytic bath. The method is specifically directed to the solidification, separation and consolidation of soluble compounds of metals including uranium and radioactive components from phosphoric acid solutions for disposal or recovery of the respective constituents. The invention is capable of dealing with phosphoric acid solutions including soluble forms of metals such as iron, nickel, molybdenum, copper, zinc, chromium, aluminum, cobalt and manganese, and radio-nuclides including plutonium, uranium, radium, cobalt, strontium, and americium. Uranium may be present in the phosphoric acid solution in both soluble and insoluble compounds comprising UO.sub.2, U.sub.3 O.sub.8, UO.sub.4, (NH.sub.4).sub.2 U.sub.2 O.sub.7, CaU.sub.2 O.sub.7, UF.sub.4 and UO.sub.2 F.sub.2. Referring to the diagram of the drawing, an electroplating system is illustrated comprising an electrolytic bath containing phosphoric acid as the electrolyte for the electropolishing of metal in accordance with the techniques of the art. This invention deals with the spent electrolyte from such a system, comprising phosphoric acid solution containing metals dissolved therein and retained as soluble compounds in the acid medium. The metal-containing acid solution effluent from the electrolytic bath system can be subjected to a preliminary treatment of its contact with an anion exchange medium when it is appropriate to increase the concentration ratio of the dissolved metals to the acid of the solution, and to reclaim phosphoric acid for return and reuse within the electrolytic bath of the system. The effluent from the electrolytic bath can be diluted if its viscosity is so high (for example a specific gravity of greater than about 1.5) as to retard effective percolation through a particulate mass of the ion exchange medium and a preliminary filtering for the purpose of excluding any solids from obstructing flow through the ion exchange material and polluting the exchange material. Contact with an anion exchange material removes a substantial portion of the phosphoric acid by preferential diffusion into the exchange material from the electrolyte solution effluent, which results in an increased proportion of dissolved metals to free phosphoric acid in solution free of the remaining exchange material. The phosphoric acid absorbed by the anion exchange material can be recovered and recycled to the electrolytic bath for reuse by washing the resin free of acid with water. The acid solution or the effluent from the ion exchange material, or unit containing same, consisting of a high dissolved metal-to-acid solution is treated for the initial metal precipitation by the addition of a solution of a carbonate of an alkali metal. The carbonate solution is introduced in an amount sufficient to lower the pH of the acid solution to about 5 to 6. Heat and agitation can be applied to the solution following the carbonate addition to expel carbon dioxide therefrom. Sodium carbonate (Na.sub.2 CO.sub.3) is preferred for this precipitation, but other useful carbonates comprise sodium bicarbonate, potassium carbonate, and potassium bicarbonate. The reaction of the preferred sodium carbonate with the phosphoric acid solution is shown in the equation: EQU 4H.sub.3 PO.sub.4 +3Na.sub.2 CO.sub.3 .fwdarw.2Na.sub.2 HPO.sub.4 +2NaH.sub.2 PO.sub.4 +3H.sub.2 O+3CO.sub.2 [ 1] Typical metals in the solution form mixed hydroxides and phosphates of generally low solubilities whereby the bulk of the initial soluble metals are precipitated out of solution at this stage of the process. However, when such a solution contains soluble uranium compounds, a portion thereof remain soluble as a tri-carbonate complex. The solids precipitated from the acid solution by the addition of the carbonate are separated and removed from the liquid portion by conventional means, such as filtration, settling or centrifuging. The remaining filtrate or supernate solution is passed to the next operation while the solids recovered are retained for a suitable disposal. When the soluble uranium compound content of the solution is significant, (for example about 5 or more parts per million by weight), it is preferred to subject the solution to an intermediate precipitation treatment of the addition thereto of sodium hydrosulfite (Na.sub.2 S.sub.2 O.sub.4) solution. The reaction of the sodium hydrosulfite in the phosphoric acid solution produces uranous bi-phosphate having the following formula: EQU U(HPO.sub.4).sub.2 To maximize the uranous precipitation produced by the addition of sodium hydrosulfite, the phosphoric acid solution should be neutralized with a carbonate solution such as sodium carbonate with an adjusted concentration which leaves the resultant neutralized supernate at the optimum phosphate concentration at about 1.1.+-.0.2 moles per liter. If the resultant phosphate concentration is too high, it can be reduced by water dilution, or if it is too low, it can be increased by evaporation. Also, to maximize the precipitation, it is preferred that the sodium hydrosulfite be introduced into the solution while at a temperature of about 75.degree. C. and with agitation. The solids precipitated from the solution by the addition of the sodium hydrosulfite are separated and recovered from the liquid portion by conventional means, including filtration, settling or centrifuging. The remaining filtrate or supernate solution is passed to the next operation while the solids recovered are retained for a suitable disposal. The final precipitation of the sequence comprises the addition of a soluble calcium salt, comprising calcium nitrate or calcium chloride, and also an alkali metal hydroxide to the filtrate or supernate solution from the former precipitation and solids separation. The preferred calcium salt is calcium nitrate, which is added first. The pH of the solution is subsequently adjusted to a basic condition with the metal hydroxide, preferably sodium hydroxide. The reaction mechanism for this precipitation is shown in the equations: EQU 2Na.sub.2 HPO.sub.4 +3Ca(NO.sub.3).sub.2 .fwdarw.Ca.sub.3 (PO.sub.4).sub.2 +4NaNO.sub.3 +2HNO.sub.3 [ 2] EQU 2HNO.sub.3 +2NaOH.fwdarw.2NaNO.sub.3 +2H.sub.2 O [3] As indicated, following the addition of calcium nitrate, a pH reversal occurs due to the formation of nitric acid. The fromed nitric acid reverses the reaction of equation 2 allowing for some phosphate solubility. The hydroxide is added to reverse this occurrence of solubility, and also to minimize the solubility of any contained residual metal ions. At a pH of about 10, the individual metal hydroxide ions' solubilities are at or near their minimums. All metals are present below 2 parts per million in the resultant solution filtrate or supernate and copper, molybdenum, cobalt, chromium and uranium are below 1 part per million. The calcium nitrate can be added at about 10 up to about 50 percent excess phosphate stoichiometry with good results. The precipitation reaction is preferably carried out in a hot solution of about 50.degree. C. The solids precipitated from the solution by the addition of the calcium salt and the alkali metal hydroxide are separated and recovered from the liquid portion by conventional means, including filtration, settling or centrifuging. The precipitated solids from each of the foregoing solidification and separation operations can be dehydrated to reduce their volume and combined if appropriate, for disposal or salvage. The filtrate or supernate solution from the foregoing sequence of solidification and separation operations, consisting primarily of a solution of sodium nitrate, and essentially free of radio-nuclides, can be safely disposed of in a waste retention lagoon or in other apt waste repositories. The procedures of the invention provides for the separation of potentially radioactive materials from a liquid medium and their solidification and reduction to a minimum volume for isolation and storage in a safe and efficient manner. |
053961412 | claims | 1. A radioisotopic power source comprising: a substrate of semiconductor material, said substrate including integrated circuitry formed therein; a trench formed in said substrate; a PN junction formed along the wall of said trench; a first power lead connected to the P portion of said PN junction; a second power lead connected to the N portion of said PN junction; and a radioactive source deposited in said trench. 2. The radioisotopic power source of claim 1, wherein said radioactive source comprises an .alpha. particle emitting radioactive source. 3. The radioisotopic power source of claim 1, wherein said radioactive source comprises a .beta. particle emitting radioactive source. 4. The radioisotopic power source of claim 1, wherein said radioactive source comprises a photon emitting radioactive source. 5. The radioisotopic power source of claim 1, wherein said radioactive source comprises a charged particle emitting radiation source. |
abstract | A vol-oxidizer of spent nuclear fuel, the spent nuclear fuel is injected to a reaction portion, the reaction portion is connected to a driving portion and oxidizes the spent nuclear fuel by rotating and back-rotating the spent nuclear fuel. The oxidized powder of the spent nuclear fuel is gathered in a discharge portion located in a lower portion of the reaction portion. By providing minute powder particles for recycling and a post process of the spent nuclear fuel, even though a size of an apparatus is small, processing a large amount is possible. Time required for oxidation can be reduced, and the powder is readily discharged by gravity since the apparatus is vertically configured. |
|
063079186 | description | DETAILED DESCRIPTION OF THE INVENTION Referring to FIGS. 1 and 2, a computed tomography (CT) imaging system 10 is shown as including a gantry 12 representative of a "third generation" CT scanner. Gantry 12 has an x-ray source 14 that projects a beam of x-rays 16 toward a detector array 18 on the opposite side of gantry 12. Detector array 18 is formed by detector elements 20 which together sense the projected x-rays that pass through a medical patient 22. Each detector element 20 produces an electrical signal that represents the intensity of an impinging x-ray beam and hence the attenuation of the beam as it passes through patient 22. During a scan to acquire x-ray projection data, gantry 12 and the components mounted thereon rotate about a center of rotation 24. Rotation of gantry 12 and the operation of x-ray source 14 are governed by a control mechanism 26 of CT system 10. Control mechanism 26 includes an x-ray controller 28 that provides power and timing signals to x-ray source 14 and a gantry motor controller 30 that controls the rotational speed and position of gantry 12. A data acquisition system (DAS) 32 in control mechanism 26 samples analog data from detector elements 20 and converts the data to digital signals for subsequent processing. An image reconstructor 34 receives sampled and digitized x-ray data from DAS 32 and performs high speed image reconstruction. The reconstructed image is applied as an input to a computer 36 which stores the image in a mass storage device 38. Computer 36 also receives and supplies signals via a user interface, or graphical user interface (GUI). Specifically, computer receives commands and scanning parameters from an operator via console 40 that has a keyboard and a mouse (not shown). An associated cathode ray tube display 42 allows the operator to observe the reconstructed image and other data from computer 36. The operator supplied commands and parameters are used by computer 36 to provide control signals and information to x-ray controller 28, gantry motor controller 30, DAS 32, and table motor controller 44. As shown in FIGS. 3 and 4, detector array 18 includes a plurality of detector modules 58. Each detector module 58 is secured to a detector housing 60. Each module 58 includes a multidimensional scintillator array 62 and a high density semiconductor array (not visible). A post patient collimator (not shown) is positioned over and adjacent scintillator array 62 to collimate x-ray beams before such beams impinge upon scintillator array 62. Scintillator array 62 includes a plurality of scintillation elements arranged in an array, and the semiconductor array includes a plurality of photodiodes (not visible) arranged in an identical array. The photodiodes are deposited, or formed on a substrate 64, and scintillator array 62 is positioned over and secured to substrate 64. Detector module 58 also includes a switch apparatus 66 electrically coupled to a decoder 68. Switch apparatus 66 is a multidimensional semiconductor switch array of similar size as the photodiode array. In one embodiment, switch apparatus 66 includes an array of field effect transistors (not shown) with each field effect transistor (FET) having an input, an output, and a control line (not shown). Switch apparatus 66 is coupled between the photodiode array and DAS 32. Particularly, each switch apparatus FET input is electrically connected to a photodiode array output and each switch apparatus FET output is electrically connected to DAS 32, for example, using flexible electrical cable 70. Decoder 68 controls the operation of switch apparatus 66 to enable, disable, or combine the outputs of the photodiode array in accordance with a desired number of slices and slice resolutions for each slice. Decoder 68, in one embodiment, is a decoder chip or a FET controller as known in the art. Decoder 68 includes a plurality of output and control lines coupled to switch apparatus 66 and computer 36. Particularly, the decoder outputs are electrically connected to the switch apparatus control lines to enable switch apparatus 66 to transmit the proper data from the switch apparatus inputs to the switch apparatus outputs. The decoder control lines are electrically connected to the switch apparatus control lines and determine which of the decoder outputs will be enabled. Utilizing decoder 68, specific FETs within switch apparatus 66 are enabled, disable, or combined so that specific outputs of the photodiode array are electrically connected to CT system DAS 32. In one embodiment defined as a 16 slice mode, decoder 68 enables switch apparatus 66 so that all rows of the photodiode array are electrically connected to DAS 32, resulting in 16 separate, simultaneous slices of data being sent to DAS 32. Of course, many other slice combinations are possible. In one specific embodiment, detector 18 includes fifty-seven detector modules 58. The semiconductor array and scintillator array 62 each have an array size of 16.times.16. As a result, detector 18 has 16 rows and 912 columns (16.times.57 modules), which enables 16 simultaneous slices of data to be collected with each rotation of gantry 12. Of course, the present invention is not limited to any specific array size, and it is contemplated that the array can be larger or smaller depending upon the specific operator needs. Also, detector 18 may be operated in many different slice thickness and number modes, e.g., one, two, and four slice modes. For example, the FETs can be configured in the four slice mode, so that data is collected for four slices from one or more rows of the photodiode array. Depending upon the specific configuration of the FETs as defined by decoder control lines, various combinations of outputs of the photodiode array can be enabled, disabled, or combined so that the slice thickness may, for example, be 1.25 mm, 2.5 mm, 3.75 mm, or 5 mm. Additional examples include a single slice mode including one slice with slices ranging from 1.25 mm thick to 20 mm thick, and a two slice mode including two slices with slices ranging from 1.25 mm thick to 10 mm thick. Additional modes beyond those described are possible. FIGS. 5 and 6 are schematic views of one embodiment of system 10 in accordance with the present invention. X-ray beam 16 emanates from a focal spot 90 of x-ray source 14. The intensity and quality of x-ray beam 16 is altered by filter assembly 92, and filtered beam 16 is projected toward detector array 18. More specifically and in one embodiment, filter assembly 92 includes a fixed filter portion 94, a z-axis movable filter 98 having a first portion 100 and a second portion 102. Respective portions 100 and 102 are configured to alter the intensity and quality of x-ray beam 16. More specifically, the shape and material composition of respective portions 100 and 102 are configured so that unique, or different, quality and intensity beams are created by filter assembly 92 based upon the position the movable filter 98. Particularly, first portion 100 includes a first filter material 110, a second filter material 112 and a third filter material 114 positioned between, or interposed, materials 110 and 112. In one embodiment, first portion 100 is configured as a bowtie filter and respective materials 110, 112, and 114 are graphite, aluminum, and copper. For example, material 110 may be 54.0 mm thick, material 112 may be 6.0mm thick, and material 114 is about 75 micrometers thick so that first portion 100 is configured to generate a harder x-ray beam quality, for example to perform a body scan. In one embodiment, third layer 114 is positioned between layers 110 and 112 so that third layer 114 is protected from damaged during operation and handling. In alternative embodiments, materials 110, 112, and 114 may selected from other attenuating materials of various thicknesses. Further, one skilled in the art could select materials 110, 112, 114 from elements, compounds, or epoxy mixtures with approximately similar mass attenuation coefficients and adjust the thickness to compensate for material density differences. For example, material 110 (54 mm of graphite) could be substituted with 95 mm of polyethylene or material 114 (75 micrometers of copper) could be substituted with 325 micrometers of titanium. Second portion 102, in one embodiment, includes a first filter material 120 and a second filter material 122. The physical configuration and selection for respective materials 120 and 122 are selected so that an x-ray beam radiating from second portion 102 has an intensity and quality unique from an x-ray beam radiating from first portion 100. In one embodiment, second portion 102 is configured to generate a softer x-ray beam quality and materials 120 and 122 are selected from the same materials as respective materials 110 and 112, however the physical shape of material 120 and 122 are altered. For example, second portion 102 is fabricated as a bowtie filter for generating a narrower x-ray beam and second portion filter material 120 is graphite and second filter material 122 is aluminum. Utilizing the described second portion 102, a head scan can be performed. In alternative embodiments, respective shape and filter materials 120 and 122 may be selected from other shapes and materials other than materials 110 and 112. In addition, similar to portion 100, second portion 102 may include any number of materials. In one embodiment, filter assembly 92 further includes a drive assembly 116 coupled to movable filter 92. Drive assembly 116 is configured to alter the z-axis position of movable filter 92 so that the intensity and quality of x-ray beam 16 can be altered. Drive assembly, in one embodiment, is coupled to computer 36. In alternative embodiments drive assembly may also be coupled to a filter drive controller (not shown). In operation, after selecting a scan type, movable filter 92 is positioned so that the proper x-ray beam is radiated toward patient 22, or an object. After collecting image data of the object or patient 22 using detector array 18, a reconstructed image is generated. More specifically, the scan type is initially determined using known selection criteria, or is prescribed by the operator, for example, a body scan. Utilizing the scan type information, movable filter 92 is positioned so that x-ray beam 16 is filtered using the appropriate portion of movable filter 92. More particularly and in one embodiment, an appropriate quality and intensity x-ray beam is generated by positioning movable filter 92 so that x-ray beam 16 is radiated into first portion 100, for example for a body scan. The z-axis position of movable filter 92 is adjusted by drive assembly 116. If another type of scan is selected, for example, a head scan, the position of movable filter 92 is adjusted, or repositioned. More specifically, using drive assembly 116, the z-axis position of movable filter 92 is adjusted so that x-ray beam 16 is filtered by second portion 102. The filtering by second portion 102 alters the x-ray beam so that, for example for the head scan, x-ray beam 16 radiated toward detector array 18 is narrower and the beam quality is softer. Utilizing portions 100 and 102, x-ray beam 16 is filtered so that the proper x-ray beam intensity and quality is directed toward detector array 18. In alternative embodiments, movable filter 92 may include any number of portions so that any number of unique quality and intensity beams may be radiated toward patient 22 and detector array 18. The above described filter assembly allows selection of filtration characteristics depending upon the scan to be completed. More specifically, the filter assembly includes a plurality of filters so that the proper filtration is provided for various specific types of scans, i.e., head portion scans or body portion scans. From the preceding description of various embodiments of the present invention, it is evident that the objects of the invention are attained. Although the invention has been described and illustrated in detail, it is to be clearly understood that the same is intended by way of illustration and example only and is not to be taken by way of limitation. For example, the CT system described herein is a "third generation" system in which both the x-ray source and detector rotate with the gantry. Many other CT systems including "fourth generation" systems wherein the detector is a full-ring stationary detector and only the x-ray source rotates with the gantry, may be used. Similarly, while the systems described herein have been two-slice and four-slice, any multi-slice system may be used. Accordingly, the spirit and scope of the invention are to be limited only by the terms of the appended claims. |
051606976 | summary | FIELD OF THE INVENTION The invention relates to a lower connector of a fuel assembly for a nuclear reactor, particularly for a pressurized-water nuclear reactor, comprising an adapter plate of square shape, traversed by water passage orifices, and a filtration means consisting of a plate pierced with holes of small dimensions and abutting against the adapter plate. BACKGROUND OF THE INVENTION Pressurized-water nuclear reactors comprise a core consisting of prism-shaped assemblies arranged side by side in vertical position. The assemblies comprise a framework which is closed by connectors and in which are arranged the fuel rods held by spacer grids spaced from one another in the longitudinal direction of the assembly. The spacer grids form a regular network, in which some locations are occupied by guide tubes intended for receiving the absorbent rods of control clusters ensuring control of the power released by the core of the nuclear reactor. At least some of the guide tubes are connected to the two connectors of the assembly by means of their end parts and ensure the junction between the framework components and the rigidity of the framework. One of the connectors of the assemblies, called the lower connector, comes to rest on the lower core plate, which is pierced with holes in the region of each of the assemblies in order to allow the cooling water of the reactor to pass through the core in the vertical direction and from the bottom upwards. The cooling fluid for the fuel rods passes through the adapter plate of the lower connector via orifices called water passages. Debris which may be present in the primary circuit of the reactor is liable to be carried along by the pressurized water in circulation, and when it is of small size (for example, smaller than 10 mm) this debris can pass through the adapter plate of the lower connector, the water passages of which have a large cross-section. This debris can become jammed between the fuel rods and the cells of the first grid, i.e., of the lowermost spacer grid holding the rods in the form of a regular network. This debris, subjected to the axial and transverse hydraulic stresses which are high in this zone, can wear the jacket of the fuel rod. This may result in a loss of sealing of this jacket and an increase in the rate of activity of the primary circuit of the reactor. Devices making it possible to filter the cooling fluid while the reactor is in operation are known in the art. Such filter elements are associated with the fuel assemblies and are generally arranged in their lower connector. They generally consist of sheet-metal or metal-wire structures making it possible to detain debris having a size smaller than the largest dimension of the passage cross-section between a fuel rod and a grid cell. Such filter elements are described, for example, in U.S. Pat. Nos. 4,664,880, 4,684,496 and EP-A-0,196,611. Such devices can be complex and introduce a relatively high head loss into the circulation of the cooling fluid through the fuel assembly. Moreover, these devices placed in the lower connector of the assembly can be bulky and obstructive during the loading and unloading of core assemblies and during the dismounting and refitting of the connections of the guide tubes and lower connector. To deal with these problems, applicants' patent application FR-A-89-04840 therefore proposed a lower connector for a fuel assembly comprising a device for the retention of particles contained in the cooling fluid of the reactor, consisting of a filtration plate pierced with holes and fastened against the lower face of the adapter plate over a substantial part of its surface. The adapter plate of such a connector comprises, in general terms, water passage orifices of cylindrical shape and of circular cross-section which extend through the adapter plate of square shape in the form of a square-mesh network. The filtration plate comprises sets of orifices of small dimensions which come into alignment with the water passage holes of the adapter plate when the filtration plate is put in place against the lower face of the adapter plate. The arrangement of the water passage orifices in the form of a regular network and the arrangement of the holes of the filtration plate in coincidence with the water passage orifices, as described in patent application FR-A-89-04840, make it possible to obtain a water passage flow in the fuel assemblies which is maintained at an acceptable level while the reactor is in operation. Nevertheless, it is still desirable to have the possibility of substantially increasing the flow of the primary fluid consisting of the pressurized water passing through the adapter plate of the lower connectors of the fuel assemblies. It is likewise desirable to make it easier to machine the adapter plate and to limit the deformations of this plate as far as possible when the connector is assembled by welding. SUMMARY OF THE INVENTION The object of the invention is, therefore, to provide a lower connector of a fuel assembly for a nuclear reactor comprising an adapter plate of square shape, traversed by water passage orifices, and a filtration means consisting of a plate pierced with holes of small dimensions and abutting against the adapter plate, this connector making it possible substantially to increase the flow of the primary fluid penetrating into the fuel assembly, the adapter plate being designed, furthermore, so as to be capable of being machined easily and, if appropriate, welded when the connector is being assembled, without experiencing deformation in a preferred direction. To this end, the water passage orifices of the adapter plate are arranged completely symmetrically in relation to the medians and to the diagonals of the adapter plate, the set of orifices of each of the zones comprises orifices having an oblong cross-section comprises orifices having a cross-section of oblong shape and, if appropriate, water passage orifices of a different shape, and the filtration plate comprises sets of holes arranged in zones of this plate coming into alignment with the oblong orifices of the adapter plate. In a preferred embodiment, each of the zones of the adapter plate delimited by a median and a diagonal comprises not only orifices of oblong shape, but also orifices of cylindrical shape and of circular cross-section, likewise arranged completely symmetrically in relation to the medians and to the diagonals of the adapter plate. |
059998946 | summary | BACKGROUND OF THE INVENTION 1. Field of the Invention The invention relates to a method for the analysis of process data of an industrial plant, in particular of a power station plant, the parts of which plant are automatically controlled. In a control room for controlling an industrial plant, in particular a power station plant, large quantities of various measurement data occur continuously and in their entirety describe a plant state or operating state. The operating personnel of the plant have the task of identifying the measurement data or measurement variables that are respectively relevant to the operating state as well as the task of following, analyzing and interpreting their values in relation to the state of the plant. At the same time, the guidance of the process from the control room is largely determined through screens by way of standards and guidelines in the form of regulations. Those regulations include symbols for parts of the plant or elements of the plant such as, for example, pumps and valves, the coloring of indicators and the construction of the indicators of a control system. In addition to the various indicators, there is commonly a plant diagram which represents the entire plant in overview. However, with increasing automation and complexity of such an industrial plant, the number of measurement data that are recorded also increases, and therefore the probability that information important to the respective operating state of the plant is not identified as such in good time. Corresponding counter measures can thus only be belatedly initiated. SUMMARY OF THE INVENTION It is accordingly an object of the invention to provide a method for the analysis of process data of an industrial plant, in particular of a power station plant, which overcomes the hereinafore-mentioned disadvantages of the heretofore-known methods of this general type, with which special features in a plant process are indicated directly and with which it is possible to take counter measures in good time, in particular in the case of disturbances. With the foregoing and other objects in view there is provided, in accordance with the invention, a method for the analysis of process data of an industrial plant, in particular a power station plant, having automatically controlled plant parts, which comprises prescribing features characterizing a plant process and providing parameters relevant to the plant process; checking the presence of each feature for each plant part using the parameters; determining correlations between combinations of plant parts or features, respectively, using features common to various plant parts and plant parts common to various features; and representing the plant parts and/or the features as information elements positioned in relation to one another, for representing a degree of correlation of two of the information elements in each case by a distance between the two information elements. In this configuration, the invention proceeds from the consideration that, on the basis of the mathematical model of formal concept analysis, large quantities of process data can be filtered, compressed and/or structured in accordance with the principle that "contextual proximity corresponds to spatial proximity" in relation to their significance for the plant state. The fundamentals of formal concept analysis are, for example, summarized in publications M. Luxemburger entitled: "Implikationen, Abhangigkeiten und Galois-Abbildungen. Beitrage zur formalen Begriffsanalyse" [Implications, Dependencies and Galois Mapping. Contributions to Formal Concept Analysis], Dissertation, TH Darmstadt (1993) and Proc. NATO Adv. Study Inst., Banff, Canada 1981, pages 445-470 (1982), as well as a publication by G. Kalmbach entitled: "Diskrete Mathematik. Ein Intensivkurs fur Studienanfanger mit Turbo-Pascal-Programmen" [Discrete Mathematics. An Intensive Course for Beginners, with Turbo-Pascal Programs] (1981). A list of features is provided from the information system of the industrial plant, which is part of the operating system of the plant. The features as a whole describe all of the possible operating states or plant states. The features themselves are, for example, status messages or other messages which describe the state of a part of the plant uniquely and which, for their part, can be members of series of messages. The assignment of the features to the parts of the plant is carried out by using parameters that are currently recorded or modeled (simulated) and which are likewise provided by the information system of the industrial plant. The contextual proximity or the proximity in terms of content of two parts of the plant in each case is then determined by the relationship of the number of the features which are common to them to the number of those features that are exhibited by at least one of the parts of the plant. In other words: in each case two parts of the plant which agree in all features are classified as particularly close in terms of content, whereas two parts of the plant which agree in none of the features are classified as not close in terms of content. In order to provide the graphical representation, the proximity in terms of content of two parts of the plant is transformed into a spatial proximity of information elements representing the parts of the plant. The spatial proximity of two features in each case is determined in an analogous manner, with use being made of the number of those parts of the plant which exhibit these features in common. The assignment of features to each part of the plant exhibiting them, using the parameters, uniquely determines the correlation or the relationship between these features and this part of the plant. In accordance with another mode of the invention, the parameters provided by the information system of the industrial plant are a component part of event messages which characterize changes of operating states or deviations from the normal state of the plant. In this case, the event messages are uniquely assigned to the corresponding parts of the plant using specific identifiers. The graphical representation which is generated can be merely a configuration of information elements representing parts of the plant or merely a configuration of information elements representing features. Preferably, however, information elements both of parts of the plant and of features are represented graphically. In accordance with a further mode of the invention, the positioning of the information elements in relation to one another within the configuration is determined in such a manner that the following criterion is fulfilled: if a part of the plant exhibits a feature, the distance of its information elements in relation to one another is smaller than a prescribable first limiting value. If a part of the plant does not exhibit a feature, the distance between its information elements is greater than a prescribable second limiting value. In order to enable the information which is important for an identification of the respective plant state to be provided to the operating personnel in a particularly simple and/or clear manner, the measurement data picked up within the plant process or parameters derived therefrom are filtered. In accordance with an added mode of the invention, a determination as to which of the parts of the plant are represented is made on the basis of a prescribed criterion. For example, those parts of the plant can be represented which agree in one feature such as, for example, in the state "disturbance"/"no disturbance", or in the status ON/OFF. In accordance with an additional mode of the invention, a time window is prescribed as a criterion, so that relationships or interactions can be detected between those parts of the plant which report disturbances within a specific time interval. As a result, conclusions can be drawn regarding causative disturbances, in contrast to symptomatic disturbances. In accordance with yet another mode of the invention, in order to be able to detect a development trend in the direction of a disturbance in good time, a time window can also be prescribed as feature. In this way, a time-based ordering of the information elements representing the parts of the plant and features is possible. In accordance with yet a further mode of the invention, information elements of successive event messages are represented together as a state complex. The state complex in this case can have a characteristic structure, the pattern of which has a direct relationship with a system behavior. In accordance with yet an added mode of the invention, in this case a system state is forecast from the common representation of the information elements. In this way it is possible to already counteract an incipient disturbance in a suitable manner during the initial stage. In accordance with yet an additional mode of the invention, the state complex is compared with a reference complex derived from plant-specific knowledge. Patterns which are characteristic of a specific plant behavior are then prescribed in this reference complex. For example, a disturbance proceeding from a rapid closure of a safety valve in a power station plant can be provided in the form of a reference complex. Comparing the state complex with this reference complex therefore permits an on-line incipient rapid closure to be detected on the basis of the disturbance messages proceeding therefrom. The information space in which the information elements are represented is n-dimensional and preferably 3-dimensional. In accordance with a concomitant mode of the invention, three spatial coordinates are determined in order to fix the position of each information element in this information space. A 3-dimensional representation is thus possible on a suitable display device, for example on a monitor screen, in the control room. 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 method for the analysis of process data of an industrial plant, 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. The construction and method of operation of the invention, however, together with additional objects and advantages thereof will be best understood from the following description of specific embodiments when read in connection with the accompanying drawings. |
summary | ||
abstract | The present disclosure provides a neutron capture therapy system. The neutron capture therapy system includes an accelerator, the accelerator generates a charged particle beam; a neutron generator, the neutron generator generates a neutron beam after being irradiated by the charged particle beam; a beam shaping assembly, the beam shaping assembly includes a moderator and a reflector surrounds around the outer periphery of the moderator, the moderator moderates the neutrons generated by the neutron generator to a preset spectrum, and the reflector leads the deflected neutrons back to increase the neutron intensity within the preset spectrum; and a collimator, the collimator concentrates the neutrons generated by the neutron generator; the spectrum of the neutron beam is changed by changing the spectrum of the charged particle beam. |
|
abstract | A CRUD collection system for removing deposits from a nuclear reactor fuel rod submerged in pool water. The system includes a sealable collection head for containing a portion of the fuel rod. The collection head includes at least one seal positioned to receive the fuel rod. Also provided is a liquid inlet communicating with an interior portion of the collection head, the inlet is configured to inject a liquid into the interior to displace the pool water therefrom. At least one blade is provided that is positionable in scraping engagement with the fuel rod. Also included is a liquid outlet for eliminating a CRUD sample and the liquid to a filter. |
|
abstract | A preventive maintenance method and apparatus for a structural member in a reactor pressure vessel according to the present invention reduce a tensile residual stress on a surface of the structural member by impinging a water jet from a nozzle onto a plane surface of a deflector to thereby change direction of flow of the water jet, and impinging the water jet after being deflected onto the surface of the structural member. This method and apparatus are applicable to a narrow space portion, and can improve a residual stress on the surface of the structural member and can also prevent damage such as stress corrosion cracking. |
|
claims | 1. An antenna characteristic measurement apparatus for measuring a characteristic of an antenna of a radio communication apparatus using a human phantom apparatus,wherein said human phantom apparatus comprises:a body section;a head section connected with said body section;at least one shoulder section connected with said body section; andan arm section including a hand section, said arm section connected with said shoulder section,wherein each of the body section, the head section, the at least one shoulder section, and the arm section is filled with a human body equivalent material,wherein the human phantom apparatus has an attitude of holding a radio communication apparatus by the hand section of said arm section, so that the human phantom apparatus looks at a display unit of the radio communication apparatus in front of said body section,wherein said antenna characteristic measurement apparatus comprises a control device for measuring the characteristic of the antenna of the radio communication apparatus, by changing at least one parameter selected from (a) an interval between said body section and the radio communication apparatus, (b) an inclined angle of the radio communication apparatus with respect to a horizontal direction, and (c) a height of the radio communication apparatus,wherein said human phantom apparatus comprises a finger phantom apparatus, andwherein said finger phantom apparatus comprises:a hollow fingertip section made of an elastic material; anda hollow finger root section made of a dielectric material,wherein said fingertip section and said finger root section are filled with a human body equivalent material, and are sealed. 2. The antenna characteristic measurement apparatus as claimed in claim 1,wherein the characteristic of the antenna of said radio communication apparatus is measured in such a state that said fingertip section is brought into contact with an input device of said radio communication apparatus. |
|
050826182 | claims | 1. Method for modifying the concentration of boric acid in the form of soluble poison contained in cooling fluid of a primary circuit of a nuclear reactor, comprising the steps of (a) removing said cooling fluid from said primary circuit (1); (b) monitoring said concentration of boric acid in said cooling fluid; (c) as a function of said concentration of said cooling fluid, selectively passing all of said cooling fluid into at least one electrodialysis module, and alternatively passing only a first part of said cooling fluid into electrodialysis modules and a second part towards at least one reverse osmosis apparatus; (d) adding an additive promoting dissociation of said boric acid to said cooling fluid when said cooling fluid leaves said primary circuit; and (e) returning said cooling fluid into said primary circuit. 2. Method according to claim 1, wherein said additive is ammonia solution. 3. Method according to claim 2, including the step of readjusting the content of ammonia solution of said cooling fluid of said primary circuit upstream of said electrodialysis modules and of said reverse osmosis apparatus. |
claims | 1. A fuel assembly, comprising:a plurality of fuel rods containing a fissile material;a lower tie plate which supports each lower end portion of the fuel rods;an upper tie plate which holds each upper end portion of the fuel rods;a plurality of fuel spacers, each of which bundles the plurality of fuel rods;a channel box attached to the upper tie plate, extending toward the lower tie plate, and surrounding the plurality of fuel rods bundled by the fuel spacers; andflow resistance members, which are disposed in an inner side of an outermost layer region of an array of the plurality of fuel rods in a cross section in a direction perpendicular to a center axis of the channel box, and through which coolant paths are formed, and which the fuel rods penetrate, whereinwhen the number of the fuel rods held by the flow resistance member is R, the number of all the fuel rods in the fuel assembly is A, and a projected area ratio C of projected area Sa of the flow resistance member from an upper tie plate side to projected area Sb of the fuel spacer which holds all the fuel rods in the fuel assembly from the upper tie plate side is defined by the following formula (1), the projected area ratio C is within a range of 1.5 to 5.2,C=(Sa/Sb)×(A/R) (1),Sa is smaller than Sb,in one of the flow resistance members, ferrules are arranged in a square grid of six rows and six columns in a square-shaped band, and two water rod disposition regions, each of which occupies a region capable of disposing four ferrules, are formed in a central portion,adjacent ferrules are joined to each other by welding, and each of the ferrules disposed in an outermost layer is joined to the square-shaped band by welding,in each of the ferrules, only two protrusions are formed in a circumferential direction, and only a single spring member is provided across adjacent pairs of ferrules and attached thereto,a bridge member is provided only at every two adjacent ferrules facing a water rod disposition region and is attached to a side surface of each ferrule, anda spring member is attached to only one bridge member among the four bridge members facing each of the water rod disposition region. 2. The fuel assembly according to claim 1, whereinwhen a distance from a center of a given ferrule to the center of a ferrule which is immediately adjacent thereto and comes into contact with the given ferrule is taken as 1,the distance from the center of the given ferrule to the center of the ferrule which is immediately adjacent thereto and comes into contact with another ferrule in a direction orthogonal to a straight line connecting the center of the given ferrule to a center of the another ferrule is √2, andthe given ferrule and the ferrule do not come into contact with each other, and a space is formed between the given ferrule and the ferrule, so that a space formed between the given ferrule and the ferrule formed between two ferrules immediately adjacent to each other in a diagonal direction of the flow resistance member,some of the resistance members are round plate-shaped members made of a zirconium alloy, and come into contact with each side surface of immediately adjacent four ferrules and are fixed by welding to the side surfaces of the immediately adjacent four ferrules,other resistance members are round plate-shaped members made of a zirconium alloy, and are disposed in each second space formed by the square-shaped band, anda diameter of the other resistance members is smaller than a diameter of the some of the resistance members. |
|
abstract | A process for chemical stabilization of a uranium carbide composite material: UCx+yC with x≧1 and y>0, placed in a stabilization chamber, comprises: a rise in chamber internal temperature for oxidation of the compound based on uranium carbide between approximately 380° C. and 550° C., the chamber being fed with a neutral gas; isothermal oxidative treatment at the oxidation temperature, the chamber being placed under O2 partial pressure; controlling completion of stabilization of the compound, comprising monitoring the amount of molecular oxygen consumed and/or carbon dioxide or carbon dioxide and carbon monoxide given off, until achievement of an input set-point value for the amount of molecular oxygen, of a minimum threshold value for the amount of carbon dioxide or minimum threshold values for the carbon dioxide and carbon monoxide. A device implements the process. |
|
description | The present application is a continuation-in-part of U.S. patent application Ser. No. 13/353,888 filed Jan. 19, 2012, and a continuation-in-part of U.S. patent application Ser. No. 13/353,923 filed Jan. 19, 2012, both of which claim priority to U.S. patent application Ser. No. 29/283,507 now Pat. No. D657,886 filed on Jan. 19, 2011. The present disclosure relates generally to nuclear pharmacy generators and tools for use therewith. Nuclear medicine uses radioactive material for diagnostic and therapeutic purposes by injecting a patient with a dose of the radioactive material, which concentrates in certain organs or biological regions of the patient. Radioactive materials typically used for nuclear medicine include Technetium-99m, Indium-111, and Thallium-201 among others. Some chemical forms of radioactive materials naturally concentrate in a particular tissue, for example, radioiodine (I-131) concentrates in the thyroid. Radioactive materials are often combined with a tagging or organ-seeking agent, which targets the radioactive material for the desired organ or biologic region of the patient. These radioactive materials alone or in combination with a tagging agent are typically referred to as radiopharmaceuticals in the field of nuclear medicine. At relatively low doses of radiation from a radiopharmaceutical, a radiation imaging system (e.g., a gamma camera) may be utilized to provide an image of the organ or biological region in which the radiopharmaceutical localizes. Irregularities in the image are often indicative of a pathology, such as cancer. Higher doses of a radiopharmaceutical may be used to deliver a therapeutic dose of radiation directly to the pathologic tissue, such as cancer cells. A variety of systems are used to generate, enclose, transport, dispense, and administer radiopharmaceuticals. One such system includes a nuclear pharmacy generator, including an elution column, and an input connector (e.g., an input needle) and an output connector (e.g., an output needle) in fluid communication with the elution column. Typically, a radiopharmacist or technician fluidly connects an eluant vial (e.g., a vial containing saline) to the input connector and fluidly connects an empty elution vial (e.g., a vial having at least a partial internal vacuum) to the output connector. The vacuum in the empty elution vial draws the eluant (e.g., saline) from the eluant vial through the elution column, and into the elution vial. The saline elutes radioisotopes as its flows through the elution column so that radioisotope-containing saline fills the elution vial. The elution vial is typically housed in its own radiation shielding container, sometimes referred to as pharmacy shield or an elution shield. Due to the use of radioactive materials, the nuclear pharmacy generator requires a radiation shield to protect the technician operating the generator. Known generator radiation shields may be comprised of a plurality of stackable painted rings that surround the generator. These rings may be comprised of any material known to block harmful radiation, such as lead. The rings may sustain considerable damage during loading and unloading of the generator and during daily elution. Damage to the radiation shielding rings may cause the painted surface to chip or crack. Moreover, during the required cleaning of the rings, a disinfecting liquid may seep into the seams between the individual rings that are stacked to form the radiation shield. After the cleaning process is completed, the liquid may leak from the seams and result in an improperly cleaned work area. This Background section is intended to introduce the reader to various aspects of art that may be related to various aspects of the present disclosure, which are described and/or claimed below. This discussion is believed to be helpful in providing the reader with background information to facilitate a better understanding of the various aspects of the present disclosure. Accordingly, it should be understood that these statements are to be read in this light, and not as admissions of prior art. A protective shroud for covering a radiation shield is provided. The radiation shield includes seams between components that comprise the radiation shield. The shroud comprises a first opening adapted for receiving the radiation shield. A second opening in the shroud allows for access into the radiation shield. The shroud also comprises a tubular body that is configured to contain the radiation shield. Referring to FIGS. 1A-4, one embodiment of a radioisotope elution system 10 includes a radioisotope generator 12 (FIGS. 3 and 4), which is removably receivable in an auxiliary shield assembly 14. As explained in more detail below, an elution tool 16, which houses an elution vial 17 (broadly, a container), and an eluant vial 18 (broadly, a container) are fluidly connectable to the radioisotope generator 12. Herein, “fluidly connectable” refers to the ability of first component and a second component to be connected (either directly or indirectly) or interface in a manner such that fluid (e.g., eluate, eluant) may flow therebetween in a substantially confined flow path. The auxiliary shield assembly 14 includes a radiation shielding body 20 that defines a cavity 22 in which the generator 12 is removably receivable, and a radiation shielding lid 24 that may be positioned on the body 20 toward a top thereof to substantially enclose the cavity 22 defined in the body 20. The auxiliary shield assembly 14 may further include an auxiliary shield cover 75 configured to receive the radiation shielding body 20 and protect the body 20 from damage. The auxiliary shield cover 75 and the auxiliary shield assembly 14 are described in further detail below. In general, the radiation shielding lid 24 facilitates proper alignment of the eluant vial 18 with the radioisotope generator 12 when fluidly connecting the eluant vial with the radioisotope generator. Additional disclosure of the radiation shielding lid 24 is set forth in detail below. The elution tool 16 illustrated in FIGS. 1-11 may be of any appropriate configuration (e.g., size, shape, design), as is known to one having ordinary skill in the art, and may include one or more suitable radiation shielding materials, such as depleted uranium, tungsten, tungsten impregnated plastic, or lead. A second embodiment of the elution tool is illustrated in FIGS. 22-33 and described in detail below. The illustrated elution vial 17 is a generally cylindrical container, made from glass or other material (e.g., plastic), which includes a septum 17a secured to a top portion thereof by a metal ring or cap 17b, as is generally known in the art. The elution vial 17 may be a different type of container suitably connectable to a radioisotope generator and/or may have a shape other than generally cylindrical. In one embodiment, the interior of the elution vial 17 is at least partially evacuated such that the elution vial has a reduced internal pressure (i.e., at least a partial vacuum). The eluant vial 18, like the elution vial 17, may be a generally cylindrical container, which includes a septum (not shown) secured to a top portion thereof by a metal ring or cap (not shown), as is generally known in the art. The eluant vial 18 may be a different type of container suitably connectable to a radioisotope generator and/or may have a shape other than generally cylindrical. The eluant vial 18 is filled with an eluant fluid, such as saline. In one embodiment, the volume of eluant fluid is less than the volume of the elution vial 17. In another embodiment, the interior volume of eluant vial 18 is less than the interior volume of the elution vial 17. For example, the eluant vial 18 may have an internal volume of about 26 milliliters, and the interior volume of the elution vial 17 may be about 36 milliliters. The elution vial 17 and/or the eluant vial 18 may be of other configurations without departing from the scope of the present disclosure. Referring to FIGS. 3-5, the radioisotope generator 12 includes: a housing 26; an elution column assembly 28 (FIG. 3) disposed within the housing; and input and output connectors 30, 32, respectively, in fluid communication with the elution column assembly 28; and a hood or cap 38 secured to the housing. The generator housing 26 is generally cylindrical and defines an axially extending cavity in which the elution column assembly 28 is received. The housing cap 38 may be snap-fit on the housing 26, or secured thereto in any other appropriate manner. The housing cap 38 has a recessed portion 40 extending downward from an upper surface of the cap. The cap 38 also has a generally U-shaped channel 42 extending downward from the upper surface and through a sidewall of the cap to the recessed portion 40. As explained in more detail below, the recessed portion 40 and the channel 42 together constitute an alignment structure, more specifically female alignment structure, for facilitating proper alignment of the radiation shielding lid 24 on the generator 12. The generator housing 26 and cap 38 may be formed from plastic (such as by molding) or from other suitable, preferably lightweight, material. Moreover, the generator housing 26 itself may be free from lead, tungsten, tungsten impregnated plastic, depleted uranium, or other radiation shielding material, such that the housing provides little or only nominal radiation shielding. The generator 12 includes a generator handle 44 pivotally secured to the cap 38. The handle 44 is pivotable between a stored position, in which the handle lies in a plane substantially transverse to the axis A1 of the housing 26 (FIG. 3) and below the upper surface of the cap 38, and a carrying position, in which the handle lies in a plane substantially parallel to the axis of the housing and above the upper surface of the cap. The generator handle 44 allows a radiopharmacist or technician to lift the generator 12 for placement of the generator in the auxiliary shield assembly 14 and removal of the generator 12 from the auxiliary shield assembly 14. Specifically, the generator handle 44 allows for the generator to be placed within the auxiliary shielding body 20. The generator handle 44 may be formed from plastic or any other appropriate material and may be pivotally connected to the generator housing 26 by pivot connectors 46 (FIG. 5) or in any other appropriate manner of connection. Referring to FIG. 3, the input and output connectors 30, 32 extend upward from the elution column assembly 28 and through respective input opening 50 and output opening 52 in a bottom surface 53 of the recessed portion 40 of the generator cap 38 such that respective terminal ends or tips 30a, 32a of the input and output connectors are disposed within the recessed portion. In the illustrated embodiment, the input and output connectors 30, 32 respectively include input and output needles or needles 30, 32 for piercing respective septums 17a of the elution vial 17 and the eluant vial 18, although it is contemplated that the connectors may be of other configurations/types. In addition to the input and output connectors 30, 32, a venting needle 54, in fluid communication with atmosphere, extends through the bottom surface 53 of the recessed portion 40 of the cap 38. The venting needle 54 is adjacent to the input connector 30 and extends through the same input opening 50 in the generator cap 38. In the illustrated embodiment, the venting needle 54 includes a needle having a terminal end or tip 54a disposed within the recessed portion 40 of the generator cap 38. The venting needle 54 pierces the septum 17a of the eluant vial 18, like the input needle 30, to vent the eluant vial 18 to atmosphere. As shown in FIGS. 12-13, in a non-use configuration of the generator—such as during shipping—the generator 12 may include needle covers 55a, 55b and a cap cover 56. In the illustrated embodiment, the needle covers include an input/venting needle cover 55a removably secured directly to the input needle 30 and the venting needle 54, and an output needle cover 55b removably secured directly to the output needle 32. The needle covers 55a, 55b protect the respective needles 30, 32, 54 and inhibit contaminants from entering the elution column assembly 28 via the needles. The illustrated needle covers 55a, 55b are solid, non-hollow, one-piece members made of a suitable material (e.g., silicone) that is pierceable by the needles 30, 32, 54. Before operating the elution system 10, a technician can remove the needle covers 55a, 55b using forceps or another suitable instrument. It is understood that the elution system 10 may not include the needles covers 55a, 55b, or the needle covers may be of other configurations without departing from the scope of the present invention. Referring still to FIGS. 12-13, the cap cover 56 is removably insertable in the recessed portion 40 of the generator cap 38 to cover and protect the input, output, and venting needles 30, 32, 54, respectively. The cap cover 56 has a top surface 56a that is disposed over and covers the needles 30, 32, 54 when the cap cover is secured to the generator 12, and a sidewall 56b depending downward from the top surface that frictionally engages the sidewall of the recessed portion 40 such that the cap cover is removably retained in the recessed portion by friction-fit connection. The cap cover 56 has two finger recesses 57 in the top surface 56a thereof, and a thumb recess 58 in the top surface and the sidewall 56b thereof. A technician can grip and remove the cap cover 56 using a single hand by inserting one or more of his/her fingers into each of the finger recesses 57 and inserting his/her thumb into the thumb recess 58, and then lifting the cap cover upward and out of the recessed portion 40. It is understood that a cap cover may have other configurations and/or can be removably secured to the generator 12 in other ways without departing from the scope of the present invention. It is also understood that the elution system 10 may not include a cap cover without departing from the scope of the present invention. Referring to FIGS. 14-17, one embodiment of the elution column assembly 28 is shown in detail. As shown in FIGS. 16 and 17, an input conduit 59 extends from the input connector 30 and into a top 60a of an elution column 60 to fluidly connect the input connector to the elution column. An output conduit 61 extends from a bottom 60b of the elution column 60 to the output connector 32 to fluidly connect the elution column to the output connector. The input and output conduits 59, 61, respectively, can be made from suitable material, such as Inconel 625. The elution column 60 may include a source of radioactive material therein (e.g., molybdenum-99, adsorbed to the surfaces of beads of alumina or a resin exchange column). In the illustrated embodiment, a filter 62 (e.g., a conventional 0.2 micron filter) is fluidly connected to, and in line with, the output conduit 61. A fillport needle 63 is fluidly connected to conduit 64, which is in turn fluidly connected to the elution column 60 for loading the product (fillport needle is typically only accessed during loading and not accessed by the technician). A cover 63a, similar to the needle covers 55a, 55b described above, is removably attached to the needle 63. A venting conduit 65 (FIG. 17) fluidly connects the venting needle 54 with the atmosphere. The venting conduit 65 has a terminal end on which an air filter 66 is secured. As shown in FIGS. 14-16, a generally rigid U-shaped support 67, which may be formed from plastic or other suitable, generally rigid material, provides structural support to the input and output needles 30, 32, the venting needle 54, and the fillport needle 63, and portions of the respective conduits 59, 61, 64, 65. As shown in FIGS. 14 and 15, the elution column assembly 28 also includes a conduit shield 68 and a column shield 69. The conduit shield 68 covers the respective conduits 59, 61, 64, 65, or portions thereof, from adjacent the input, output, and venting needles, 30, 32, 54, respectively, to adjacent the top 60a of the elution column 60b. The conduit shield 68 also covers the fillport needle 63 and the output filter 62. The conduit shield 68 defines internal passages for receiving and covering the respective components, while leaving the input, output, and venting needles 30, 32, 54 and the air filter 66 exposed. The conduit shield 68 may be a two-piece construction and may include (e.g., be made from or have in their construct) lead, tungsten, tungsten impregnated plastic, depleted uranium and/or another suitable radiation shielding material. Referring to FIGS. 14 AND 15, the column shield 69 defines a chamber (not shown) for receiving the elution column 60 and a lower portion 71 of the conduit shield 68 therein. The column shield 64 may be a one-piece construction and may include (e.g., be made from or have in their construct) lead, tungsten, tungsten impregnated plastic, depleted uranium and/or another suitable radiation shielding material. Referring back to FIGS. 1A-1C, the illustrated auxiliary shield assembly body 20 includes a top ring 72, a base ring 73, and a plurality of step-shaped or generally tiered, modular rings 74, which are disposed one over the other between the base ring 73 and the top ring 72 such that a seam 300 is defined between adjacent rings. Substantially all or part of the illustrated auxiliary shield assembly body 20 may be made of one or more suitable radiation shielding materials, such as depleted uranium, tungsten, tungsten impregnated plastic, or lead. In the exemplary embodiment, rings 72, 73, and 74 are comprised of lead. In this embodiment, the top ring 72 and the modular rings 74 are substantially identical (the top ring is modular), and the top ring 72 is identified as such for ease of description only. FIG. 1A shows the shield body 20 comprising seven rings 74, however, the shield body 20 may comprise any number of rings 74 as is required to cover the generator 12. The modular aspect of the rings 74 may tend to enhance adjustment of the height of the auxiliary shield assembly body 20, and the step-shaped configuration may tend to contain some radiation that might otherwise escape through seams 300. In one embodiment (FIG. 1B), an auxiliary shield cover 75 is receivable over the shield body 20. FIG. 1B illustrates the shield cover 75 partially enshrouding the radiation shield body 20. The shield cover 75 has a smooth exterior surface and includes a first opening 304 for receiving the shield body 20 and a second opening 302 for allowing access to the radiation shielding lid 24 and the radioisotope, or nuclear pharmacy, generator 12 housed within the auxiliary shield assembly 14, and more specifically, housed within the shield body 20. The shield cover 75 further includes a tubular body 306 extending between the first and second openings 304 and 302 that is configured to contain the shield body 20 within the shield cover 75. In certain embodiments, the shield cover 75 includes at least one flange 308 that is configured to retain the shield body 20 within the cover 75. The flange 308 is perpendicular to the tubular body 306 and forms the second opening 302 such that the flange 308 surrounds the second opening 302. The flange 308 extends a distance over the top ring 74 inward between the tubular body 306 and the second opening 302, thus the second opening 302 has a diameter less than that of the first opening 304. Flange 308 extends inward from the tubular body 306 perpendicular to the tubular body 306. In the embodiment of FIG. 1C, the shield cover 75 is configured to act as a shroud and slide over the shield body 20 such that the first opening 304 first receives the top ring 72 and is advanced down the shield body 20 until substantially all of the base ring 73 is covered by the shield cover 75. The second opening 302 is configured such that the shield body 20 does not pass through the second opening 302 and the cover rests atop the shield body. That is, the diameter of the second opening 302 is less than the diameter of the radiation shield body 20. Specifically, where the flange 308 is configured to conform to the top of the shield body 20, thus retaining the shield body 20, and more specifically, the top ring 72, the second opening 302 allows for access into the shield body 20 and specifically to the radiation shielding lid 24 and the elution tool 16. The shield cover 75 is a protective shroud that is configured to provide protection from exposed lead of the rings 72, 73, and 74 that comprise the shield body 20 to a technician that is operating the radioisotope generator 12. It is known that exposure to lead may cause harmful health effects, so by covering the lead shield body 20, the shield cover 75 provides a barrier between lead, which may be exposed due to damage to the rings, and a human technician. The shield cover 75 covers the outer surface of the shield body 20, thus covering any previously exposed lead and preventing the further damage to the rings of the shield body 20 from loading/unloading of the generator 12. The shield cover 75 may be constructed in any manner from any known suitable material consistent with providing the protection described. Accordingly, in certain embodiments, a suitable material, may be one or more of a foamed or vulcanized rubber, neoprene, polyurethane, plastics, silicone, and/or a silicone containing material, though other materials are contemplated. In certain embodiments, the shield cover 75 has a smooth exterior surface that substantially covers every seam 300 between adjacent rings of the shield body 20. The continuously smooth surface provides for easy cleaning and does not allow liquid to seep into the seams 300 and potentially leak out at a later time. The shield cover 75 has a smooth surface that allows for a clean and disinfected work area because fluid is not able to leak through the shield body 20. In certain embodiments, the shield cover 75 may also include features for the storage of various tools used during the elution process. For example, a raised boss 310 may be disposed on the flange 308 of the shield cover 75 proximate to the second opening 302. The shield cover 75 may also include at least one ring or hoop 312, at least a portion of which rests on the flange 308 and extends parallel to the flange 308 over the tubular body 306. In one embodiment, the hoop 312 may be a wire metal hoop. The boss 310 is configured to temporarily store eluant shield 136 (shown in FIG. 18) when eluant shield 136 is not in use on the generator 12. The hoop 312 is configured to temporarily store vial holder 250 (shown in FIG. 34) when the generator 12 is not in use. A second hoop 313 may be used to store the recovering tool 290 shown in FIG. 38) when the generator 12 is not in use. The second hoop 313 may be of a different size than hoop 312 to receive a different sized tool. The shield cover 75 may include more than one hoop 312 such that both the vial holder 250 and the recovering tool 290 may be temporarily stored simultaneously. The boss 310 and the at least one hoop 312 serve to more efficiently position the eluant shield 136, the vial holder 250, and the recovering tool 290 for immediate access to the technician and to minimize the chance that the eluant shield 136, the vial holder 250, and the recovering tool 290 are damaged from dropping or contaminated from coming into contact with an unsanitary surface. Referring now to FIGS. 6-11, the radiation shielding lid 24 includes: a generally cylindrical lid body 76 having upper and lower surfaces, 77, 78, respectively; an elution tool opening 79; and an eluant vial opening 80. In one example (of which an exemplary method of making is explained in more detail below), the lid body 76 includes a radiation shielding core 124 that is overmolded with a plastic material 126, 128. As an example, the radiation shielding core 124 may include depleted uranium, tungsten, tungsten impregnated plastic, or lead. The upper and lower surfaces 77, 78, respectively, are generally planar, although the surfaces may be other than generally planar. A male alignment structure, generally indicated at 81, is provided on the lower surface 78 of the lid body 76 to facilitate proper alignment of the lid 24 on the generator 12. More specifically, the male alignment structure 81 has a shape generally corresponding with the combined shape of the recessed portion 40 and the channel 42 of the generator 12 (together, these recessed portion 40 and the channel 42 constitute a female alignment structure) so that the male alignment structure mates with the generator in order to align the elution tool opening 79 with the output needle 32 and the eluant vial opening 80 with the input needle 30 and the venting needle 54. As such, it may be said that the lid 24 is keyed with the generator 12 (e.g., the cap 38 thereof) such that proper positioning of the lid 24 atop the generator 12 results in alignment of the respective openings 79, 80 with the corresponding needles 32, 30. The structure 81 enables only one position of the lid 24 relative to the generator 12. The illustrated male alignment structure 81 includes a wall 81a projecting outward from the bottom surface 78 and surrounding the elution tool opening 79 and the eluant vial opening 80. A plurality (e.g., a pair) of handles 82 on the upper surface 77 of the lid body 76 allows the radiopharmacist or technician to properly place the lid 24 on the generator 12 and remove the lid from the generator. The elution tool opening 79 extends through the lid body 76 from the upper surface 77 through the lower surface 78 thereof. The elution tool opening 79 is sized and shaped for removably receiving the elution tool 16 therein. For example, in the illustrated embodiment, the elution tool opening 79 has a generally circular circumference that is substantially uniform along its axis. In one embodiment, the elution tool opening 79 has a diameter slightly larger than an outer diameter of the elution tool 16 such that the opening effectively aligns the septum (not shown) of the elution vial 17 (FIG. 4) with the output needle 32 as the elution tool is inserted into the opening. For example, the elution tool opening 79 may have a diameter that is from about 0.25 mm (0.01 in) to about 1.0 mm (0.04 in) larger than the outer diameter of the elution tool 16. In one embodiment, the elution tool opening 79 may have a diameter from about 46 mm (1.8 in) to about 48 mm (1.9 in), although it may alternatively have a diameter falling outside this range. Other shapes and sizes of the elution tool opening 79 may be appropriate; however, it tends to be preferred that the shape and size of the elution tool opening 79 be at least generally complimentary to the shape and size of the elution tool 16 being used with the radiation shielding lid 24 to reduce the likelihood of misalignment between the elution vial 17 and the output needle 32. As shown in FIGS. 9 and 10, the eluant vial opening 80 is spaced apart and separate from the elution tool opening 79, and is sized and shaped for removably receiving an eluant vial 18 (FIG. 2), such as a vial containing saline or other eluants. In the illustrated embodiment (FIG. 10), the eluant vial opening 80 has a lower end 86 at the lower surface 78 of the lid body 76 and an upper end 88 intermediate the upper and lower surfaces 77, 78, respectively. In one example, the eluant vial opening 80 may have a diameter from about 34.0 mm (1.34 in) to about 34.5 mm (1.36 in), although it may alternatively have a diameter falling outside this range. As with the elution tool opening 79, other shapes and sizes of the eluant vial opening 80 may be appropriate; however, it tends to be preferred that the shape and size of the eluant vial opening 80 be at least generally complimentary to the shape and size of the eluant vial 18 being used with the radiation shielding lid 24 to reduce the likelihood of misalignment between the eluant vial 18 and the input needle 30 and venting needle 54. Referring to FIGS. 2, 6, 8, and 11, the illustrated lid 24 has two finger recesses 90 formed in the upper surface 77 of the lid body 76, which are diametrically opposite one another with respect to the eluant vial opening 80. The finger recesses 90 are defined by respective recessed surfaces extending downward from the upper surface 77 of the lid body 76 to the eluant vial opening 80, and are sized and shaped to allow at least distal portions of two fingers of a radiopharmacist or other appropriate technician to enter the finger recesses. Recessed surfaces defining illustrated finger recesses 90 are curved and generally in the shape of a half-bowl such that the recessed surfaces lead the radiopharmacist's or technician's fingers toward the eluant vial opening 80. It is understood that in other embodiments the lid 24 may have a single finger recess, such as a finger recess that completely or partially surrounds the eluant vial opening 80, or more than two finger recesses. Referring to FIG. 8, each illustrated finger recess 90 has an upper edge 92 adjacent the upper surface 77 of the lid body 76 and a lower edge 93 that is coextensive with a portion of the upper end 88 of the eluant vial opening 80. Referring to FIG. 11, the lid 24 of the auxiliary shield assembly 14 includes first and second alignment wings 100, each designated generally at reference numeral 100, extending upward from adjacent the upper end 88 of the eluant vial opening 80 within the finger recesses 90. Each of the first and second wings 100 has opposite sides 104, a top portion 106, and an inner surface 108 extending partially around a circumference of the upper end 88 of the eluant vial opening 80. In the illustrated embodiment, the top portion 106 of each of the wings 100 is disposed above the upper surface 77 of the lid body 76 (as seen best in FIGS. 7 and 10), and the inner surface 108 of each of the wings 100 is generally arcuate, although it is understood that the wings 100 may be of other shapes and relative dimensions. Together, the inner surfaces 108 of the wings 100 and the eluant vial opening 80 define a vial passageway 107 extending from the top portions 106 of the wings 100 through the lower surface 78 of the lid body 76. The wings 100 preferably enable alignment of the eluant vial septum with the input needle 30 and venting needle 54 as the eluant vial 18 is inserted into the vial passageway 107. As such, the wings 100 preferably make it is less likely that the input needle 30 or venting needle 54 will contact the metal ring or other hard part of the vial and damage the needle. In one example, the inner surface 108 of each wing 100 may extend at least 45 degrees and less than 180 degrees around the circumference of the upper end 88 of the eluant vial opening 80. In other examples, the inner surface 108 of each wing 100 may extend at least 60 degrees, or at least 90 degrees, and less than 180 degrees around the circumference of the upper end 88 of the eluant vial opening 80. Other configurations of the wings 100 do not depart from the scope of the present disclosure. To facilitate gripping of the eluant vial 18 during at least one of insertion of the vial into the vial passageway 107 and removal of the vial from the vial passageway, the respective adjacent sides 104 of the first and second wings 100 are spaced apart from one another about the eluant vial opening 80 to define gaps or first and second finger channels, each indicated at 112 (FIGS. 6 and 10), leading from the finger recesses 90 to the vial passageway. In the illustrated embodiment, the finger channels 112 are diametrically aligned, relative to the vial opening 80, with the finger recesses 90, and the respective sides 104 of the wings 100 extend into the associated finger recesses 90. Each of the first and second finger channels 112 are sized and shaped to allow at least the distal portion of one of the two fingers to enter the corresponding finger channel from the associated finger recess 90. For example, a minimum width of each of the finger channels 112 (i.e., the distance between the respective adjacent sides 104 of the first and second wings 100) may measure from about 19 mm (0.75 in) to about 21 mm (0.83 in), and more specifically, from about 19.0 mm (0.748 in) to about 19.6 mm (0.776 in), although the minimum width of each finger channel may fall outside this range. Thus, the finger channels 112 allow the radiopharmacist or technician to grip the eluant vial 18, such as by using his/her thumb and forefinger, during at least one of insertion of the vial in the vial passageway 107 and removal of the vial from the vial passageway. In the illustrated embodiment (FIGS. 8, 10, and 11), a diameter of a portion of the vial passageway 107 defined by the inner surfaces 108 of the wings 100 tapers from the top portions 106 of the wings toward the eluant vial opening 80. Tapering the inner surfaces 108 of the wings 100 facilitates molding of the wings when overmolding the lid 24 in one example, as described below. Although this diameter of the vial passageway 107, as defined by the inner surfaces 108, tapers along the length of the passageway, a plurality of alignment ribs 114 are provided on the inner surfaces to define an effective inner diameter of the vial passageway that is substantially uniform along the length of the passageway. The ribs 114 are spaced apart from one another between the sides 104 of the wings and extend longitudinally along the respective wings 100. The wings 100 project inwardly, generally toward a centerline of the passageway 107, such that each rib 114 has a terminal, guiding surface 115 (FIG. 11) generally facing the centerline of the passageway. Each guiding surface 115 is uniformly spaced from the centerline of the vial passageway 107 along its length. In other words, the guiding surface 115 of each rib 114 does not taper or flare with respect to the axis of the vial passageway 107. Through this configuration, the guiding surfaces 115 effectively align the elution vial 18 with the input needle 30 and venting needle 54 even though the inner surfaces 108 of the wings 100 are tapered. The ribs 114 have depths projecting into the vial passageway 107 relative to the respective inner surfaces 108. Because the diameter of the vial passageway 107 defined by the inner surfaces 108 of the wings 100 tapers, yet the guiding surfaces 115 do not taper or flare relative to the centerline of the vial passageway, the depths of the ribs relative to the respective inner surfaces 108 taper toward the eluant vial opening 80. The wings 100 may not include the ribs 114 without departing from the scope of the present disclosure. As illustrated in FIG. 3, a bottom 116 of the eluant vial 18 lies slightly below or at the top portions 106 of the wings 100 when the eluant vial is received in the vial passageway 107 and fluidly connected to the input needle 30. Notches 118 in the top portions 106 of the wings 100 allow the radiopharmacist or technician to view the eluant vial 18 in the passageway without having to position his/her head above the upper surface 77 of the lid 24. In one example, the auxiliary shield lid 24 may be formed by a two-step overmolding process. In such a process, a radiation shielding core 124 (FIG. 10)—which may include a suitable radiation shielding material such as depleted uranium, tungsten, tungsten impregnated plastic, or lead—is provided. The core 124 may be generally disk-shaped, having first and second openings, which will form the elution tool and eluant vial openings, 79, 80, respectively, and recesses, which will form the finger recesses 90. A first molded part is molded with a first thermoplastic material 126 to form the bottom surface 78, the male alignment structure 81, and the sidewall of the body 76, and at least lower portions of the elution tool opening 79 and the eluant vial opening 80. Next, the core 124 is placed into the first molded part. Finally, this assembly is overmolded with a second thermoplastic material 128 to form the top surface 77, the handles 82, the finger recesses 90, the wings 100, and an upper portion of at least the elution tool opening 79. The first and second thermoplastic materials 126, 128, respectively, may include polypropylene and polycarbonate, or other material, and the first and second thermoplastic materials may be of the same material. Other methods of making the auxiliary shield lid 24 may be used. Referring to FIGS. 18-21, an eluant shield 136 of the elution system 10 is positionable over the eluant vial 18 when the vial is received in the eluant vial opening 80 in the lid 24 and fluidly connected to the generator 12 to inhibit exposure of the radiopharmacist or technician to radiation when the eluant is fluidly connected to the generator (e.g., during and after an elution process). The eluant shield 136 has a top 138, an opposing bottom 140, and a cavity 142 extending from the bottom toward the top. A pair of shielding wings 144 at the bottom 140 of the eluant shield 136 partially surround the cavity 142. The shielding wings 144 are sized and shaped to fit snugly within the finger recesses 90 in the lid 24 so that the top portions 106 of the alignment wings 100 are received in the cavity 142 of the eluant shield 136 and the shielding wings 144 oppose the sides 104 of the alignment wings and the finger channels or gaps 112 between the sides of the alignment wings. As such, substantially an entirety of the eluant vial 18 is surrounded by radiation shielding material of either the lid 24 or the eluant shield 136. More specifically, when the eluant shield 136 is positioned on the lid 24, substantially the entirety of the eluant vial 18 is surrounded by a suitable radiation shielding material, such as depleted uranium, tungsten, tungsten impregnated plastic, or lead. In one example, the eluant shield 136 may be formed by a two-step overmolding process. In such a process, a radiation shielding core 124, which may include a suitable radiation shielding material such as depleted uranium, tungsten, tungsten impregnated plastic, or lead—is provided. The core is substantially the same shape as the eluant shield in finished form, including a pair of shielding wings and a cavity. A first molded part is molded with a first thermoplastic material to form the top 138. Next, the core is placed into the first molded part. Finally, this assembly is overmolded with a second thermoplastic material to form the bottom 140, the shielding wings 144, and the cavity 142. The first and second thermoplastic materials, respectively, may include polypropylene and polycarbonate, or other material, and the first and second thermoplastic materials may be of the same material. Other methods of making the eluant shield 136 may be used. Referring to FIGS. 22-33, a second embodiment of an elution tool 150 is generally indicated at reference numeral 150. This elution tool 150 includes a body, generally indicated at 152, having a top 154, and opposing bottom 156; and a lid, generally indicated at 158, hingedly secured to the top of the elution tool body. As explained in more detail below, a dispensing cap 160 (FIG. 22) is removably securable to the bottom 156 of the elution tool body 152 for configuring the elution tool in a dispensing tool configuration, and a storage cap 162 (FIG. 23) is removably securable to the bottom of the elution tool body for configuring the elution tool into a storage tool configuration. In generally, the dispensing cap 160 and the storage cap 162 are interchangeably securable to the elution tool body 152. In the illustrated embodiment, neither the dispensing cap 160 nor the storage cap 162 are secured to the elution tool body 152 when then elution tool 150 is inserted in the auxiliary shield and the elution vial 17 in the elution tool is fluidly connected to the generator 12. The elution tool body 152 is sized and shaped to be slidably receivable in the elution tool opening 79 in the auxiliary shield lid 24. The body 152 has an upper longitudinal portion 163 having first outer diameter that defines an annular stop surface 164 to inhibit the top 154 of the body from entering the elution tool opening 79 in the auxiliary shield lid 24. A lower longitudinal portion 166 of the body 152, having a second outer diameter that is less than the first outer diameter, is receivable in the dispensing and shielding caps 160, 162, respectively, as explained in more detail below. An intermediate longitudinal portion 168 of the body 152, having an outer diameter that is less than the first outer diameter and greater than the second outer diameter OD2, is sized and shaped to be slidably receivable in the elution tool opening 79. The elution tool body 152 may include (e.g., be made from or have in their construct) lead, tungsten, tungsten impregnated plastic, depleted uranium and/or another suitable radiation shielding material. The elution tool body 152 is configured to receive the elution vial 17 therein. In particular, the elution tool body 152 has a vial chamber 170 (FIG. 33) defined therein extending from an opening 172 in the top 154 of the elution tool body to an opposing access opening 174 in the bottom thereof. The top opening 172 is sized and shaped to allow the elution vial 17 to be inserted into and removed from the vial chamber 170, and the vial chamber has a size and shape generally corresponding to the size and shape of the elution vial such that the elution vial fits generally snugly within the chamber. The bottom 156 of the elution tool body 152 defines an annular internal surface 178 surrounding the access opening 174. When the elution vial 17 is received in the vial chamber 170, the metal ring 17b of the vial contacts the internal surface 176 so that the septum 17a is aligned with the access opening 174. Accordingly, when the elution tool 150 is inserted into the elution tool opening 79 in the lid 24, the output needle 32 enters the access opening 174 and pierces the septum 17a. The elution tool lid 158 is hingedly secured to the elution tool body 152 and configurable between an open or exposed position (FIG. 24), in which the top opening 172 is exposed and the elution vial 17 can be inserted into and removed from the vial chamber 170, and a closed or occluded position (FIGS. 25-28), in which the top opening is occluded and the elution vial is retained in the vial chamber. The elution tool lid 158 includes a generally planar or disk-shaped lid body 178 that is receivable in a lid recess 180 defined in the top 154 of the elution tool body 152 when the lid is in the closed position. The lid body 178 has a lower face 178a that seats on an inner annular flange or lid seat 182 of the lid recess 180, and an upper face 178b that is substantially coplanar with the top 154 of the elution tool body 152 when the lid 158 is in a closed position. The upper face 178b of the lid body 178 has a plurality of gripping slots 179 formed therein to provide a gripping region for the radiopharmacist or technician when opening and closing the lid, as explained in more detail below. For reasons which are apparent from the below description, the elution tool lid 158 has a generally circular periphery, and the lid recess 180 and the lid seat 182 have generally oblong peripheries. Moreover, the elution tool lid 158 is sized and shaped to allow for movement of the lid along the major axis of the lid recess 180 when the lid is seated on the lid seat 182. The elution tool lid body 178, may include (e.g., be made from or have in their construct) lead, tungsten, tungsten impregnated plastic, depleted uranium and/or another suitable radiation shielding material. Referring to FIGS. 22-28, the illustrated elution tool 150 includes a hinged lid connection, generally indicated 186, and a latching mechanism, generally indicated at 188, for releasable locking the lid 158 in the closed position. The hinged lid connection 186 includes a hinge connector 190 extending radially or laterally outward from the periphery of the lid body 178, and a hinge pin 192, adjacent the periphery of the top 154 of the elution tool body 152, to which the hinge connector is coupled. The hinge connector 190 defines a slot 194 in which the hinge pin 192 is received to allow both rotation of the hinge connector (and the lid 158) about the hinge pin, and limited transverse, linear movement of the hinge connector (and the lid) relative to the hinge pin. The latching mechanism 188 includes a latching member 194 extending radially or laterally outward from the periphery of the lid body 178, generally diametrically opposite the hinge connector 190. The latching member 194 includes a tongue 196 that is slidably receivable in a latching groove 198 adjacent the periphery of the top 154 of the elution tool body 152. A detent 200 (e.g., a ball detent) on the elution tool body 152 extends into the latching groove 198 and releasably engages the latching member 194 (e.g., an underside of the latching member) as the tongue 196 is slid into the latching groove to inhibit the latching member from inadvertently withdrawing (e.g., sliding back out) from the latching groove. To lock the lid 158 in the closed position (FIGS. 27 and 28), the radiopharmacist or technician can rotate the lid about the hinge pin 192 to the closed position such that the lid body 178 is seated on the lid seat 182 of the elution tool body 152. Once seated, the slot 194 in the hinge connector 190 allows the radiopharmacist or technician to move the lid 158 linearly toward the latching groove 198, whereby the tongue 196 can be slid into the latching groove 198. For example, while holding the elution tool 150 using one hand, the radiopharmacist or technician may contact the upper face 178b of the lid body 178 (more specifically, the region defined by the gripping slots 179) with his/her thumb to rotate the lid 158 to its closed position and then linearly slide the lid toward the latching groove 198. As the latching member 194 slides over the ball detent 200, the ball detent deflects and pushes against the latching member. Once the tongue 196 is received in the latching groove 198, the lid 158 is releasably locked in the closed position. The lid 158 may be unlocked (FIGS. 25 and 26) by the radiopharmacist or technician using his/her thumb to slide the lid away from the latching groove 198, against the pushing force of the ball detent, so that the tongue 196 is withdrawn from the latching groove 198. Once unlocked, the lid 158 can be rotated to the open position. It is understood that the lid 158 may be releasably lockable in the closed position in other ways, and other ways of retaining the elution vial 17 in the elution tool 150 do not depart from the scope of the present disclosure. As disclosed above, dispensing cap 160 is removably securable to the lower longitudinal portion 168 of the elution tool body 152, such as shown in FIG. 22, to configure the elution tool in the dispensing configuration. In the dispensing configuration, the elution tool 150 can be used as a dispensing tool, whereby the radiopharmacist or technician can hold the elution tool and withdrawal a quantity of radiopharmaceutical from the elution vial 17 housed in the elution tool without removing the dispensing cap 160. The dispensing cap 160 includes a body 204 (e.g., a generally cylindrical body) having a top 206 and a bottom 208. The dispensing cap body 204 defines a socket 210 extending from the top 206 toward the bottom 208 thereof that is sized and shape for receiving the lower longitudinal portion 166 of the elution tool body 152. The socket 210 has an open top end to allow insertion of the lower longitudinal portion 166 of the elution tool body 152 into the socket, and an access opening 212 at the bottom 208 of the dispensing cap body 204 that is alignable with the access opening 174 in the elution tool body 152 to provide access to the septum 17b of the elution vial 17 in the chamber 170 of the elution tool body 152. Referring to FIG. 22, the dispensing cap 160 includes a plurality of magnetic couplers 214 attached to dispensing cap body 204 and surrounding the socket 210 for releasably securing the dispensing cap to the elution tool body 152 when the lower longitudinal portion 166 of the elution tool body is received in the socket. The magnetic couplers 214 are magnetically attracted to an annular coupler surface 216 of the elution tool body 152 that is in opposing relationship with the magnetic couplers when the lower longitudinal portion 166 of the elution tool body is received in the socket 210 of the dispensing cap 160. In another embodiment, the elution tool body 152 may include magnetic couplers that are magnetically attracted to the magnetic couplers (or some other component or structure) of the dispensing cap body 204. The dispensing cap 160 also includes a locking pin 218 extending longitudinally outward from the top 206 of the dispensing cap body 204. The locking pin 218 is alignable with and receivable in a locking cavity 220 in the annular coupler surface 216 of the elution tool body 152 to inhibit the dispensing cap 160 from rotating about the elution tool body. In one example of securing the dispensing cap 160 to the elution tool body 152, the radiopharmacist or technician may insert the lower longitudinal portion 166 of the elution tool body 152 into the socket 210 of the dispensing cap 160 and then rotate the dispensing cap about the elution tool body (or vice versa) until the locking pin 218 aligns with and enters the locking cavity 220. The dispensing cap 160 may be removably securable to the elution tool body 152 in other ways. The dispensing cap 160 includes a dispensing lid 222 pivotably secured to the bottom 208 of the dispensing cap body 204 by a pivot pin 223 (e.g., a pivot bolt) for selectively opening and closing the access opening 212 of the socket 210 and for providing suitable radiation shielding when the elution vial 17 is received in the elution tool 150. More specifically, the dispensing lid 222 is received in a recess 224 formed in the bottom 208 of the dispensing cap body 204, and is pivotable about a pivot axis defined by the pivot pin 223 that is generally parallel to the longitudinal axis of the elution tool 150. The dispensing lid 222 is pivotable between a non-dispensing position (FIGS. 29 and 30), in which the dispensing lid is aligned with and opposing (i.e., covering) the access opening 212 of the socket 210, and a dispensing position (FIGS. 30 and 31), in which the dispensing lid is at least partially misaligned with the access opening (i.e., the access opening is at least partially uncovered) to allow access to the septum 17b of the elution vial 17. A detent 226 (e.g., a ball detent) on the bottom 208 of the dispensing cap body 204 releasable locks the dispensing lid 222 in the non-dispensing position. Moreover, when the dispensing lid 222 is moved to the dispensing position, the detent 226 is removably receivable in one of a plurality of slots (e.g., three slots, not shown) formed on an underside of the dispensing lid. Accordingly, the dispensing lid 222 is releasably lockable in a selected one of a plurality of dispensing positions, each providing a different degree to which the lid is open. To position the dispensing lid 222 in the dispensing position and provide access to the elution vial 17 in the elution tool 150 when the dispensing cap 160 is secured to the elution tool, a radiopharmacist or technician can hold the elution tool in one hand and use his/her thumb to grip the dispensing lid and swing (i.e., rotate) the dispensing lid about the pivot pin 223 and away from the access opening 212 in the dispensing cap. As the radiopharmacist or technician swings the dispensing lid 222 open, the detent 226 resiliently deflects to allow the dispensing lid to slide over the detent. The radiopharmacist or technician may continue to rotate the dispensing lid 222 until the lid is at a selected dispensing position and the detent 226 enters one of the slots (not shown) on the underside of the lid. With the dispensing lid 222 in a selected dispensing position, the radiopharmaceutical in the elution vial 17 is accessible to the radiopharmacist or technician, in that the radiopharmacist or technician can insert a dispensing needle of a syringe (not shown) through the access openings 212, 174 in the respective dispensing cap 160 and the elution tool body 150 and into the elution vial 17, by piercing the septum 17b, to withdraw a desired quantity of radiopharmaceutical from the elution vial. After withdrawing the desired quantity of radiopharmaceutical, the radiopharmacist or technician can position the dispensing lid 222 in the non-dispensing position by rotating or swinging the lid toward the access opening 212, whereby the detent 226 deflects as the lid slides toward the access opening. A wall 228 partially defining the recess 224 in the dispensing cap 160 acts as a stop for inhibiting the lid from sliding past the access opening 212 as the lid being closed. The dispensing lid 222 may include (e.g., be made from or have in their construct) lead, tungsten, tungsten impregnated plastic, depleted uranium and/or another suitable radiation shielding material. In contrast, the dispensing cap body 204 may be formed from a suitable material, such as aluminum, plastic or other corrosion-resistant, lightweight material, or other material that has a density less than the density of suitable radiation shielding, such as that provided by lead, tungsten, tungsten impregnated plastic, depleted uranium. The dispensing cap body 204 does not need to provide suitable radiation shielding, such as that provided by lead, tungsten, tungsten impregnated plastic, depleted uranium and/or another suitable radiation shielding material, because such suitable radiation shielding is provided by the elution tool body 152. Accordingly, the dispensing cap 160 does not add a significant amount of weight to the elution tool 150 so that the elution tool may be suitably used as a dispensing tool for the radiopharmacist or technician. Referring to FIG. 23, as disclosed above the storage cap 162 is removably securable to the elution tool body 152 to configure the elution tool in the storage configuration. In the storage configuration, the storage cap 162 must be removed from the elution tool body 152 in order for a radiopharmacist or technician to withdraw a quantity of radiopharmaceutical from the elution vial 17. The storage cap 162 includes a storage cap body 232 (e.g., a generally cylindrical body) having a top 234 and a bottom 236, and a radiation shield 238 secured to the bottom of the storage cap body. The storage cap body 232 defines a socket 240 extending from the top 234 toward the bottom 236 of the storage cap body that is sized and shape for receiving the lower longitudinal portion 166 of the elution tool body 152. The socket 240 has an open top end to allow insertion of the lower longitudinal portion 166 of the elution tool body 152 into the socket. The radiation shield 238 is secured to the bottom 236 of the storage cap body 232 such that the shield is aligned and in opposing relationship with the access opening 174 in the elution tool body 152 when the storage cap 162 is removably secured to the elution tool 150. In the illustrated embodiment, the radiation shield 238 is a press insert into the storage cap body 232. The radiation shield 238 may be secured to the storage cap body 232 in other ways without departing from the scope of the present disclosure. Referring to FIG. 23, the storage cap 162 is removably securable to the elution tool body 152 in substantially the same way as the dispensing cap 160, although the storage cap can be removably securable in other ways. More specifically, the storage cap 162 includes a plurality of magnetic couplers 244 secured to the storage cap body 232 and surrounding the socket 240. The magnetic couplers 244 are magnetically attracted to the annular coupler surface 216 of the elution tool body 152. It is understood that the elution tool body 152 may include magnetic couplers secured thereto, that are magnetically attracted to the magnetic couplers (or another component or structure) of the storage cap body. The dispensing cap 160 may be removably securable to the elution tool body 152 in other ways without departing from the scope of the present disclosure. Referring to FIGS. 34-37, the radioisotope elution system 10 may also include a sterile vial holder, generally indicated at 250, for a vial 252 of sterile fluid (e.g., TechneStat™) in which the output needle 32 is stored when the elution system 10 is not in use. As explained in more detail below, after the elution process, the elution tool 150 may be withdrawn from the elution tool opening 79 in the auxiliary shield lid 24, at which time the sterile vial holder 250 can be inserted into the elution tool opening so that the output needle 32 pierces a septum 252a of the sterile fluid vial. The sterile vial holder 250 includes a body, generally indicated at 254, for holding the sterile vial 252 therein, and a cap, generally indicated at 256, that is removably securable to the body. The holder body 254 has a generally cylindrical receptacle 258 having an open top 260, a bottom 262, and a vial chamber 264 sized and shaped for receiving and retaining the sterile vial 252 therein. As shown in FIG. 36, the bottom 262 of the receptacle 258 defines an access opening 266 that is aligned with the septum 252a of the sterile vial 252 when the vial is received in the chamber 264 so that the output needle 32 pierces the septum and enters the sterile vial when the sterile vial holder 250 is inserted into the elution tool opening 79. The holder body 254 includes a plurality of fins 268 (e.g., four fins) projecting radially outward from the receptacle 258 and spaced apart around the receptacle. The fins 268 define a diameter or cross-sectional dimension of the receptacle 258 that is sized and shaped to fit snugly within the elution tool opening 79 so that the access opening 266 (and the septum 252a) align with the output needle 32 when the holder 250 is inserted into the elution tool opening. The holder body 254 may be of other configurations without departing from the scope of the present disclosure. The cap 256 of the sterile vial holder 250 is removably securable to the body 254 by a twist-lock mechanism, generally indicated at 270. The body 254 includes an annular female twist-lock component 272 that receives a male twist-lock component 274 projecting outward from a bottom surface 276 of the cap 256. The female twist-lock component 272 defines slots or grooves 278 that are spaced apart around an interior surface 280 of the female twist-lock component to define gaps 281. The male twist-lock component 274 includes a plurality of tabs 282 that are receivable in the gaps 281 defined between the grooves 278 of the female twist-lock component, and that enter the grooves 278 when the cap 256 is rotated about its longitudinal axis relative to the holder body 254. When the tabs 282 are received in the grooves 278, the twist-lock mechanism inhibits relative longitudinal movement between the cap 256 and the holder body 254. In the illustrated embodiment, the male twist-lock component 274 also includes a longitudinal projection 284 that enters the vial chamber 264 of the receptacle 258 and abuts the bottom of the sterile vial 252 to limit or restrict longitudinal movement of the sterile vial in the chamber. It is understood that the cap 256 may be releasably securable to the body 254 in other ways without departing from the scope of the present disclosure. The holder body 254 may be a one-piece component formed (e.g., molded) from plastic or other material that has a density less than the density of material that provides suitable radiation shielding, such as that provided by lead, tungsten, tungsten impregnated plastic, depleted uranium. The cap 256, on the other hand, may include suitable radiation shielding material such as depleted uranium, tungsten, tungsten impregnated plastic, or lead. In one example, the cap may be formed by a two-step overmolding process. In such a process, a radiation shielding core—which may include a suitable radiation shielding material such as depleted uranium, tungsten, tungsten impregnated plastic, or lead—is provided. A first molded part is molded with a first thermoplastic material to form the top 260. Next, the core is placed into the first molded part. Finally, this assembly is overmolded with a second thermoplastic material to form the bottom 262, the male twist-lock component 274, and the longitudinal projection 284. The first and second thermoplastic materials, respectively, may include polypropylene and polycarbonate, or other material, and the first and second thermoplastic materials may be of the same material. Other methods of making the cap 256 may be used. Referring to FIGS. 38 and 39, the elution system 10 may also include a re-covering tool, generally indicated at 290, for reapplying the input/venting needle cover 55a and the output needle cover 55b on the respective input and venting needles 30, 54 and the output needle 32. The re-covering tool 290 has a first longitudinal portion 292, defining an output needle cover cavity 294 for snugly receiving the output needle cover 55b therein, and a second longitudinal portion 296, defining an input/venting needle cover cavity 298 for snugly receiving the input/venting needle cover 55a therein. The first longitudinal portion 292 has a size and shape such that it is snugly receivable in the elution tool opening 79 in the auxiliary shield lid 24, and the second longitudinal portion 296 has a size and shape such that it is snugly receivable in the eluant vial opening 80 in the auxiliary shield lid. The re-covering tool 290 may be formed from plastic, or other suitable material, and may be molded as a single, one-piece structure. To reapply the covers 55a, 55b, the radiopharmacist or technician inserts the covers into the respective cavities 294, 298. The covers 55a, 55b are held in the respective cavities 294, 298 by friction-fit engagement between the walls of the cavities and the covers. The radiopharmacist or technician can then insert the second longitudinal portion 296 into the eluant vial opening 80, whereupon the input and venting needles 30, 54 pierce the cover 55a. Upon withdrawing the second longitudinal portion 296 from the eluant vial opening 80, the cover 55a remains secured to the input and venting needles 30, 54. The radiopharmacist or technician can then insert the first longitudinal portion 292 into the elution tool opening 79 to reapply the cover 55b in a similar manner. It is understood that the covers 55a, 55b may be reapplied in any order without departing from the scope of the present disclosure. In a method of using the radioisotope elution system 10, the radiopharmacist or technician manually inserts the radioisotope generator 12 into the cavity 22 of the auxiliary shield body 20, the handle is folded down, and the cap cover 56 is removed in the manner set forth above. The auxiliary shield lid 24 is then manually placed in the cavity, on top of the radioisotope generator 12. The lid 24 may be rotated to thereby mate the male alignment structure 81 on the lid with the female alignment structure (i.e., the recessed portion 40 and the U-shaped channel 42) in the cap 38 of the generator 12. Upon mating, the eluant vial opening 80 is disposed over and generally vertically aligned with the input needle 30 and the venting needle 54, and elution tool opening 79 is disposed over and generally vertically aligned with the output needle 32. Using forceps (or another tool), the radiopharmacist or technician removes the two covers 55a and 55b. The eluant vial 17 is manually inserted into the passageway defined by the wings 100 and the eluant vial opening 80. The passageway guides the eluant vial 17 in a substantially vertical direction, such that the longitudinal axis of the eluant vial is generally aligned with the axes of the input needle 30 and the venting needle 54. More specifically, the passageway guides the eluant vial 17 such that the input needle 30 and the venting needle 54 pierce the septum of the vial to fluidly connect the interior of the eluant vial to the generator 12. The radiopharmacist or technician can view the bottom 116 of the eluant vial 18 through the notches 118 in the respective wings 100 when the vial is received in the passageway 107 to confirm that the eluant vial 18 is fully inserted onto the generator 12. Accordingly, the radiopharmacist or technician does not have to position his/her head directly above the lid 24 to confirm that the needles 30, 54 actually pierced the eluant vial septum. To this effect, the radiopharmacist or technician reduces any likelihood of radiation exposure from the generator 12 when positioning his/her head over the eluant vial opening 80. Once confirmation is made that the vial is properly placed, the eluant shield 136 may be placed over the bottom of the eluant vial in the manner set forth above. In this method, the elution vial 17 is inserted into the elution tool 150 and the lid 158 is closed in the manner set forth above. The elution tool, which does not have either the dispensing cap 160 or the storage cap 162 secured thereto, is manually inserted into the elution tool opening 79 such that the output needle 32 pierces the septum of the elution vial to fluidly connect the elution vial to the generator 12. The vacuum (or reduced pressure) in the elution vial 17 draws the saline from the vial 18 through the radioisotope column and into the elution vial 17. After the elution vial 17 is filled with the desired quantity of radioisotope-containing saline, the elution tool 150 can be manually removed from the lid 24, at which time the dispensing cap 160 or the storage cap 162 can be secured to the elution tool body 152 in the manner set forth above. With the dispensing cap 160 secured to the elution tool body 152, the radiopharmacist or technician can withdraw desired quantities of the radiopharmaceutical from the elution vial 17 in the manner set forth above. With the elution tool 150 removed from the lid 24, the sterile vial holder 250 can be inserted into the elution tool opening 79 so that the output needle 32 pierces the sterile vial 252. The now empty eluant vial 18 may remain on the radioisotope generator 12 until a subsequent elution in order to keep the needles 30, 54 sterile. When it is time for a subsequent elution, the eluant vial 18 can be manually removed from lid 24, such as by the radiopharmacist or technician inserting his/her thumb and forefinger into the respective finger recesses 90 and then into the respective finger channels 112 to grip (or pinch) the eluant vial. The radiopharmacist or technician can then lift the eluant vial 18 upward off the needles 30 and 54 and out of the lid 24. When introducing elements of the present invention or the embodiment(s) thereof, the articles “a”, “an”, “the” and “said” are intended to mean that there are one or more of the elements. The terms “comprising”, “including” and “having” are intended to be inclusive and mean that there may be additional elements other than the listed elements. As various changes could be made in the above apparatus and methods without departing from the scope of the disclosure, it is intended that all matter contained in the above description and shown in the accompanying figures shall be interpreted as illustrative and not in a limiting sense. |
|
053316746 | claims | 1. In a nozzle dam for hermetically sealing against a wall of a nozzle of the steam generator bowl of a water cooled nuclear reactor primary coolant system against passage of fluids between said nozzle and said bowl, said nozzle dam comprising: passage means completely through said nozzle dam for releasing gas from the nozzle in the region below the nozzle dam, through the dam, to said bowl when said nozzle dam is mounted sealingly in said nozzle. said passage means comprising a hole through said dam, and said dam further comprising valve means connected to said hole for controlling flow of fluids through said hole in said dam. pipe means connected to said valve means, said pipe means being positioned for directing said fluids out of said bowl. after the reactor coolant system water general level is raised to a level that is higher than the nozzle dam, and before completion of a later step of lowering of the reactor coolant system water general level to bring it to a level that is below the nozzle dam wherein taking of the general level below the nozzle dam is not yet attained, the intermediate step of opening a passage exiting the nozzle for release of trapped gas through said passage from the nozzle from the region immediately below the nozzle dam. said intermediate step being carried out before said later step of lowering the reactor coolant system water general level is started. said passage for said release of said gas from said nozzle being through said nozzle dam. said passage for said release of said gas from said nozzle being through a wall of said bowl. said passage for said release of said gas from said nozzle being through said nozzle dam. said opening of said passage being by opening a valve in a pipe line of said passage until gas flow ceases. said opening of said passage being by opening a valve in a pipe line of said passage until gas flow ceases. in preparation for removing said nozzle dam from said nozzle, before completion of the step of lowering of the reactor coolant system water general level to bring it to a level that is below the nozzle dam, wherein taking of the general level below the nozzle dam is not yet attained, the step of opening a passage exiting the nozzle for release of trapped gas from the nozzle in the region immediately below the nozzle dam, and keeping the passage open until the trapped gas is released and water in the nozzle rises to the dam. 2. The nozzle dam described in claim 1, further comprising: 3. The nozzle dam described in claim 2, further comprising 4. In a water cooled nuclear reactor primary coolant system in which a nozzle dam is in place for hermetically sealing a nozzle of the steam generator bowl, a method for reducing the chance of uncontrolled drop of reactor coolant system inventory level below mid loop level when a hermetic seal between the nozzle and steam generator bowl is broken, said method comprising: 5. The method described in claim 4, further comprising: 6. The method described in claim 4, further comprising: 7. The method described in claim 4 further comprising: 8. The method described in claim 5, further comprising: 9. The method described in claim 5 further comprising said passage for said release of said gas from said nozzle being through a wall of said bowl. 10. The method described in claim 6 further comprising: 11. The method described in claim 7, further comprising: 12. In a water cooled nuclear reactor primary coolant system in which a nozzle dam is in place for hermetically sealing a nozzle of the steam generator bowl, a method for reducing the chance of uncontrolled drop of reactor coolant system inventory level below mid loop level when a hermetic seal between the nozzle and steam generator bowl is broken, said method comprising: |
045270686 | claims | 1. A concrete shielding housing for receiving and storing a transportable fuel element container which is suitable for storage and filled with spent nuclear reactor fuel elements, the housing comprising: a pallet-like base having a base plate; the housing having clear interior dimensions somewhat larger than the outer dimensions of the container; air inlet means formed in the lower portion of said housing for admitting air into the interior thereof; air outlet means formed in the upper portion of said housing for conducting air from said interior to the ambient; bore means formed in said base plate; and collection means arranged beneath said base plate and communicating with said bore means for receiving and collecting moisture formed in said interior of said housing. 2. The concrete shielding housing of claim 1, said collection means being a collection pan disposed beneath said base plate and mounted to said base. 3. The concrete shielding housing of claim 2, said collection pan being configured so as to have a location of lowest elevation, and a drain plug threadably engaging said pan at said location. 4. The concrete shielding housing of claim 1, said bore means being a plurality of bores formed in said base plate for conducting away moisture formed in said interior; and, said collection means communicating with said plurality of bores for receiving and collecting the moisture formed in said interior of said housing. |
summary | ||
061730262 | claims | 1. A power oscillation monitoring system for detection and indication of instability of a nuclear reactor having a plurality of fuel bundles and a plurality of neutron flux detectors for generating a plurality of respective output signals indicative of thermal-hydraulic oscillation frequencies of the fuel bundles, said system comprising a processor programmed to: (a) determine an instability threshold confirmation count based on the neutron flux detector output signals utilizing a period-based algorithm, determining the instability threshold confirmation count comprises generating an instability threshold flag S.sup.i.sub.th determined in accordance with the following function: ##EQU5## (b) determine a confirmation density based on the instability threshold confirmation count utilizing a confirmation density stability algorithm; and (c) generate a thermal-hydraulic instability signal when the confirmation density reaches a predetermined level. where M is the number of active neutron flux detectors in the core and S.sub.th.sup.i is the instability threshold flag. where E is an axial loss of period based algorithm efficiency, and F.sup.max.sub.CD is a bounding maximum fraction wherein the denominator is the number of active neutron detector output signals and the numerator is the number of active detector output signals that equal or exceed the instability threshold confirmation count. (a) determining an instability threshold confirmation count based on the neutron flux detector output signals utilizing a period-based algorithm; (b) determining a confirmation density based on the instability threshold confirmation count utilizing a confirmation density stability algorithm, determining the confirmation density comprises the step of generating an instability threshold flag status S.sup.i.sub.th for each of the neutron flux detector output signals in accordance with the following function: ##EQU7## (c) generating a thermal-hydraulic instability signal when the confirmation density reaches a predetermined level. where M is the number of active neutron flux detectors in the core and S.sup.i.sub.th is the instability threshold flag status for each of the neutron flux detectors. where E is an axial loss of period based algorithm efficiency, and F.sup.max.sub.CD is a bounding maximum fraction wherein the denominator is the number of active neutron detector output signals and the numerator is the number of active detector output signals that equal or exceed the instability threshold confirmation count. 2. A power oscillation monitoring system in accordance with claim 1 wherein the period-based algorithm is tuned based on a successive confirmation count model. 3. A power oscillation monitoring system in accordance with claim 1 wherein the neutron flux detectors are local power range monitors. 4. A power oscillation monitoring system in accordance with claim 1 wherein the confirmation density is a fraction wherein a denominator is the number of active neutron detector output signals and a numerator is the number of active neutron flux detector output signals that exceed a target successive oscillation period confirmation count. 5. A power oscillation monitoring system in accordance with claim 1 wherein the instability signal has a first state and a second state, and the signal remains in the first state until the confirmation density exceeds the a confirmation density setpoint. 6. A power oscillation monitoring system in accordance with claim 1 wherein the confirmation density is determined in accordance with the following function: ##EQU6## 7. A power oscillation monitoring system in accordance with claim 1 wherein the instability threshold confirmation count N.sub.th is equal to 11. 8. A power oscillation monitoring system in accordance with claim 1 wherein a maximum confirmation density setpoint is determined in accordance with the following function: EQU S.sup.max.sub.CD =(1-E) * F.sub.CD.sup.max 9. A power oscillation monitoring system in accordance with claim 8 wherein the axial loss of period based algorithm efficiency is equal to 0.25. 10. A power oscillation monitoring system in accordance with claim 8 wherein the maximum confirmation density setpoint is in the range of approximately 0.2 to 0.33. 11. A method for indicating instability of a nuclear reactor using a power oscillation monitoring system, the reactor having a core including a plurality of fuel bundles and a plurality of neutron flux detectors, the flux detectors distributed throughout the core contiguous the fuel bundles, each of the neutron detector providing an output signal being indicative of thermal-hydraulic oscillation frequencies of the fuel bundles, the system having a processor, said method comprising the steps of: 12. A method in accordance with claim 11, wherein the periodic based algorithm is tuned based on a successive confirmation count model. 13. A method in accordance with claim 11, wherein the step of generating a confirmation density is determined in accordance with the following function: ##EQU8## 14. A method in accordance with claim 11, wherein generating an instability signal further comprises the step of comparing the confirmation density to a predetermined level in accordance with the following function:. EQU CD>=S.sub.CD.sup.max 15. A method in accordance with claim 11, wherein the maximum confirmation density setpoint S.sub.CD.sup.max is in the range 0.2 to 0.33. 16. A method in accordance with claim 11 wherein the instability threshold confirmation count N.sub.th is equal to 11. 17. A method in accordance with claim 11 wherein determining a confirmation density further comprises the step of determining a maximum confirmation density setpoint in accordance with the following function: EQU S.sup.max.sub.CD =(1-E) * F.sub.CD.sup.max 18. A method in accordance with claim 17 wherein the axial loss of period based algorithm efficiency is equal to 0.25. |
abstract | A fault monitoring method for a robot (10) or other work machine (30) is disclosed, in which the period that starts when the robot or the work machine starts moving and ends when the work machine stops moving is set as one unit, the difference between the maximum torque and the minimum torque within the unit is set as a maximum fluctuation range, a fluctuation range control value is set on the basis of the maximum fluctuation range, an average is taken of the maximum fluctuation ranges at plural points during operation, a second average value (t1, t10, t100) is determined, and a fault is determined to have occurred when the second average value exceeds the fluctuation range control value. |
|
claims | 1. A radiation area monitor device, comprising:a radiation sensor having a directional radiation sensing capability;a rotation mechanism operable for selectively rotating the radiation sensor such that the directional radiation sensing capability selectively sweeps an area of interest; anda processor operable for analyzing and storing a radiation fingerprint acquired by the radiation sensor as the directional radiation sensing capability selectively sweeps the area of interest;wherein the radiation sensor comprises one or more of a gamma sensor and a neutron sensor. 2. The radiation area monitor device of claim 1, wherein the radiation sensor comprises a dual gamma/neutron radiation sensor. 3. The radiation area monitor device of claim 1, wherein the radiation area monitor device is operable for selectively operating in:a first supervised mode during which a baseline radiation fingerprint is acquired by the radiation sensor as the directional radiation sensing capability selectively sweeps the area of interest; anda second unsupervised mode during which a subsequent radiation fingerprint is acquired by the radiation sensor as the directional radiation sensing capability selectively sweeps the area of interest, wherein the subsequent radiation fingerprint is compared to the baseline radiation fingerprint and, if a predetermined difference threshold is exceeded, an alert is issued. 4. A radiation area monitor method, comprising:providing a radiation sensor having a directional radiation sensing capability;providing a rotation mechanism operable for selectively rotating the radiation sensor such that the directional radiation sensing capability selectively sweeps an area of interest; andanalyzing and storing a radiation fingerprint acquired by the radiation sensor as the directional radiation sensing capability selectively sweeps the area of interest;wherein the radiation sensor comprises one or more of a gamma sensor and a neutron sensor. 5. The radiation area monitor method of claim 4, wherein the radiation sensor comprises a dual gamma/neutron radiation sensor. 6. The radiation area monitor method of claim 4, wherein the radiation area monitor device is operable for selectively operating in:a first supervised mode during which a baseline radiation fingerprint is acquired by the radiation sensor as the directional radiation sensing capability selectively sweeps the area of interest; anda second unsupervised mode during which a subsequent radiation fingerprint is acquired by the radiation sensor as the directional radiation sensing capability selectively sweeps the area of interest, wherein the subsequent radiation fingerprint is compared to the baseline radiation fingerprint and, if a predetermined difference threshold is exceeded, an alert is issued. |
|
abstract | A method for electroplating a nonmetallic grating including providing a nonmetallic grating; performing an atomic layer deposition (ALD) reaction to form a seed layer on the nonmetallic grating; and electroplating a metallic layer on the seed layer such that the metallic layer uniformly and conformally coats the nonmetallic grating. An apparatus including a silicon substrate having gratings with an aspect-ratio of at least 20:1; a atomic layer deposition (ALD) seed layer formed on the gratings; and an electroplated metallic layer formed on the seed layer, wherein the electroplated metallic layer uniformly and conformally coats the gratings. |
|
claims | 1. A beam energy identification system, comprising:an electrostatic beam scanner configured to receive an ion beam, wherein the ion beam is deflected along a fast scan axis via a variation of a scanner voltage applied to the electrostatic beam scanner, therein defining a scanned ion beam;a first Faraday cup positioned along the fast scan axis; anda second Faraday cup positioned along the fast scan axis, wherein the scanner voltage is measured when a peak of the scanned ion beam is present at the respective first Faraday cup and second Faraday cup, and wherein a beam energy (E) is defined by:E=(Δθ1Sq)/(K1SΔV1S), wherein:Δθ1S a shift of angle of the scanned ion beam;q is a charge value of ions of the scanned ion beam;K1S is a constant throughout ranges of beam energy and scanner voltage; andΔV1s is a difference in scanner voltages associated with the respective peaks of the scanned ion beam at the first Faraday cup and the second Faraday cup. 2. A beam energy identification system, comprising:an electromagnetic beam scanner configured to receive an ion beam, wherein the ion beam is deflected along a fast scan axis via a variation of a scanner current applied to the electromagnetic beam scanner, therein defining a scanned ion beam;a first Faraday cup positioned along the fast scan axis; anda second Faraday cup positioned along the fast scan axis, wherein the scanner current is measured when a peak of the scanned ion beam is present at the respective first Faraday cup and second Faraday cup, and wherein a beam energy (E) is defined by:E=(K1MΔB1MqΔθ1M)2/m, wherein:K1M is constant throughout ranges of beam energy and scanner current;ΔB1M is a difference in a scanner magnetic field associated with the respective peaks of the scanned ion beam at the first Faraday cup and the second Faraday cup;q is a charge value of ions of the scanned ion beam;Δθ1M is a shift of angle of the scanned ion beam; andm is a mass of the ions of the scanned ion beam. 3. The beam energy identification system of claim 2, further comprising an angle corrector magnet positioned downstream of the electromagnetic beam scanner, wherein the angle corrector magnet is configured to selectively parallelize the scanned ion beam. 4. A beam energy identification system, comprising:an electrostatic beam scanner configured to receive an ion beam, wherein the ion beam is deflected along a fast scan axis via a variation of a scanner voltage applied to the electrostatic beam scanner, therein defining a scanned ion beam;an angle corrector magnet positioned downstream of the electrostatic beam scanner, wherein the angle corrector magnet is configured to parallelize the scanned ion beam, therein defining a parallel shifted ion beam;a first Faraday cup positioned along the fast scan axis; anda second Faraday cup positioned along the fast scan axis, wherein the scanner voltage is measured when a peak of the parallel shifted ion beam is present at the respective first Faraday cup and second Faraday cup, and wherein a beam energy (E) is defined by:E=(Δθ2Sq)/(f2SK2XΔV2S), wherein:Δθ2S is a shift of angle of the parallel shifted ion beam;q is a charge value of ions of the parallel shifted ion beam;f2S is a correction factor to account for an effect of the angle corrector magnet;K2S is a constant throughout ranges of the beam energy and scanner voltage; andΔV2S is a difference in scanner voltages associated with the respective peaks of the parallel shifted ion beam at the first Faraday cup and the second Faraday cup. 5. A beam energy identification system, comprising:an electromagnetic beam scanner configured to receive an ion beam, wherein the ion beam is deflected along a fast scan axis via a variation of a scanner current applied to the electromagnetic beam scanner, therein defining a scanned ion beam;an angle corrector magnet positioned downstream of the electromagnetic beam scanner, wherein the angle corrector magnet is configured to parallelize the scanned ion beam, therein defining a parallel shifted ion beam;a first Faraday cup positioned along the fast scan axis; anda second Faraday cup positioned along the fast scan axis, wherein the scanner voltage is measured when a peak of the parallel shifted ion beam is present at the respective first Faraday cup and second Faraday cup, and wherein a beam energy (E) is defined by:E=(K2MΔB2MqΔθ2M)2/m, wherein:K2M is constant throughout ranges of the beam energy and scanner current;ΔB2M is a difference in a scanner magnetic field associated with the respective peaks of the parallel shifted ion beam at the first Faraday cup and the second Faraday cup;q is a charge value of ions of the parallel shifted ion beam;Δθ2M is a shift of angle of the parallel shifted ion beam; andm is a mass of the ions of the parallel shifted ion beam. 6. A method for identifying an energy of an ion beam, the method comprising:scanning the ion beam along a fast scan axis via an application of a scanner voltage to an ion beam scanner, therein defining a scanned ion beam;positioning a first faraday cup downstream of the ion beam along the fast scan axis;positioning a second faraday cup downstream of the ion beam scanner along the fast scan axis;determining a shift of angle of the scanned ion beam, wherein the shift angle is associated with a position of the ion beam scanner relative to the first Faraday cup and second Faraday cup;varying the scanner voltage and determining the scanner voltage when a first peak and a second peak of the scanned ion beam is present at the respective first Faraday cup and second Faraday cup; anddetermining a beam energy (E) of the scanned ion beam, wherein the beam energy is defined by:E=(Δθ1Sq)/(K1SΔV1S), wherein:Δθe1S is the shift of angle of the scanned ion beam;q is a charge value of ions of the scanned ion beam;K1S is a constant throughout ranges of beam energy and scanner voltage; andΔV1S is a difference in scanner voltages associated with the respective first and second peaks of the scanned ion beam at the first Faraday cup and the second Faraday cup. 7. A method for identifying an energy of an ion beam, the method comprising:scanning the ion beam along a fast scan axis via an application of a scanner voltage to an ion beam scanner, therein defining a scanned ion beam;positioning a first Faraday cup downstream of the ion beam scanner along the fast scan axis;positioning a second Faraday cup downstream of the ion beam scanner along the fast scan axis;activating an angle corrector magnet positioned downstream of the ion beam scanner, wherein the angle corrector magnet parallelizes the scanned ion beam, therein defining a parallel shifted ion beam;determining a shift of angle of the parallel shifted ion beam, wherein the shift angle is associated with a position of the ion beam scanner relative to the first Faraday cup and second Faraday cup;varying the scanner voltage and determining the scanner voltage when a first peak and a second peak of the parallel shifted ion beam is present at the respective first Faraday cup and second Faraday cup; anddetermining a beam energy (E) of the parallel shifted ion beam, wherein the beam energy is defined by:E=(Δθ2q)/(K2ΔV2S), wherein:Δθ2 is the shift of angle of the parallel shifted ion beam;q is a charge value of ions of the parallel shifted ion beam;K2 is approximately constant throughout ranges of beam energy and scanner voltage; andΔV2S is a difference in scanner voltages associated with the respective first and second peaks of the parallel shifted ion beam at the first Faraday cup and the second Faraday cup. 8. The beam energy identification system of claim 1, further comprising an angle corrector magnet positioned downstream of the electrostatic beam scanner, wherein the angle corrector magnet is configured to selectively parallelize the scanned ion beam. |
|
summary | ||
047012986 | claims | 1. An underground nuclear reactor comprising: a cylindrical pressure vessel; a pile of spherical fuel elements located within a removable core vessel arranged within a cavity defined by said cylindrical pressure vessel; an outer side reflector laterally disposed against an external surface of said removable core vessel; an outer bottom reflector located beneath said removable core vessel; an inner bottom reflector beneath said pile within said removable core vessel and removable therewith; an inner side reflector surrounding said pile within the removable core vessel and removable therewith; a plurality of channels in said inner side reflector; absorber means for trimming and shutdown of said reactor displaceably disposed within said channels; a roof reflector located within the removable core vessel and resting directly on said fuel elements; said cylindrical pressure vessel having an opening in a cover region of sufficient size to allow removal of said removable pressure vessel; a cover arranged in said opening; blower means for forcing primary loop circulation of cooling gas downward through said pile arranged in said cover; cooling system means for transferring heat generated in said pile from the pressure vessel arranged on an inner surface of said pressure vessel; jacket means for gas tight sealing of said cooling system means against said cavity; an annular space between said outer reflector and said jacket means for transport of said cooling gas; means for structurally sealing said primary loop circulation and avoiding need of forced ventilation and filter devices; wherein said primary loop and said cooling system means operate to safely remove decay heat following an accident. foundation means for supporting said cylindrical pressure vessel; means for covering said pressure vessel resting on said foundation means; and a light construction hall enclosing said means for covering for housing auxiliary and supply systems. decay heat is transferred by thermal conduction and radiation from said outer side reflector to said cooling system means upon a pressure release accident; and decay heat is transferred by conduction from said pressure vessel to surrounding soil and to the atmosphere upon a failure of said cooling system means. 2. A nuclear reactor as in claim 1, wherein said primary loop exhibits a pressure higher than a pressure of medium in said cooling system means. 3. A nuclear reactor as in claim 1, wherein said fuel elements exhibit a high heavy metal content. 4. A nuclear reactor as in claim 1, further comprising drive means for said absorber means located in passages in said cover. 5. A nuclear reactor as in claim 1, further comprising gas conduction jacket means for separating suction and compression sides of said blower connected to said removable core vessel and located in a free space between said roof reflector and said cover. 6. A nuclear reactor as in claim 1, wherein said jacket means exhibits ribs on a side facing said annular space. 7. A nuclear reactor as in claim 6, wherein said ribs are longitudinally arranged, extend to rest on said outer side reflector, and define cooling channels. 8. A nuclear reactor as in claim 1, further comprising means for supporting said removable core vessel resting directly on a bottom surface of said pressure vessel. 9. A nuclear reactor as in claim 1, further comprising a passage for housing a blower motor in said cover, wherein said passage is equipped with a removable closure. 10. A nuclear reactor as in claim 1, further comprising: 11. A nuclear reactor as in claim 1, further comprising means for regulating reactor output by blower controlling and secondary flow of the cooling system means rates and utilizing negative temperature coefficients. 12. A nuclear reactor as in claim 11, wherein said absorber means and burnable neutron poisons initially absorb excess reactivity and said absorber means are manually displaceable to compensate for variations of excess reactivity occurring during operation of said reactor. 13. A nuclear reactor as in claim 1, further comprising a shielding bell for placing directly onto the pressure vessel and means for drawing said removable core vessel into the shielding bell following raising of said cover for discharge of said nuclear reactor following burnoff of said fuel elements. 14. A nuclear reactor as in claim 1, wherein the decay heat is removed by natural convection and a potential pressure rise in said primary loop is compensated by an overflow of the cooling gas into gas reservoirs upon failure of the blower means; |
047568747 | claims | 1. In the method for inhibiting the deposition of radioactive cobalt in a water-bearing vessel of a water-cooled nuclear reactor which comprises adding zinc ion to water entering said water-bearing vessel, the improvement in which said zinc ion contains a substantially lower proportion of the .sup.64 Zn isotope than naturally occurring zinc. 2. A method in accordance with claim 1 in which the proportion of .sup.64 Zn in said zinc ion is less than about 10%. 3. A method in accordance with claim 1 in which the proportion of .sup.64 Zn in said zinc ion is less than about 1%. 4. A method in accordance with claim 1 in which said zinc ion is substantially devoid of .sup.64 Zn. 5. A method in accordance with claim 1 in which said zinc ion is added by dissolving zinc oxide in said water. 6. A method in accordance with claim 5 in which the amount of zinc oxide dissolved is selected to achieve a concentration of about 1 to about 1,000 ppb zinc in said water-bearing vessel. 7. A method in accordance with claim 5 in which the amount of zinc oxide dissolved is selected to achieve a concentration of about 3 to about 100 ppb zinc in said water-bearing vessel. 8. A method in accordance with claim 5 in which said zinc oxide is dissolved from a member selected from the group consisting of a water-based slurry, a water-based paste and a water-based solution. 9. A method for inhibiting the deposition of radioactive material in a water-bearing vessel of a water-cooled nuclear reactor, said method comprising: adding a water-based zinc oxide paste to water entering said water-bearing vessel, the zinc in said paste containing a substantially lower proportion of the .sup.64 Zn isotope than naturally occurring zinc. adding an aqueous solution of zinc oxide to water entering said water-bearing vessel, the zinc in said zinc oxide containing a substantially lower proportion of the .sup.64 Zn isotope than naturally occurring zinc. adding a slurry of zinc oxide in water to water entering said water-bearing vessel, the zinc in said zinc oxide containing a substantially lower proportion of the .sup.64 Zn isotope than naturally occurring zinc. 10. A method in accordance with claim 9 in which the amount of zinc oxide in said paste is from about 25% to about 95% by weight. 11. A method in accordance with claim 9 in which the amount of zinc oxide in said paste is from about 40% to about 80% by weight. 12. A method for inhibiting the deposition of radioactive cobalt in a water-bearing vessel of a water-cooled nuclear reactor, said method comprising: 13. A method in accordance with claim 12 in which said aqueous solution is formed by passing a stream of water drawn from said water entering said water-bearing vessel through a receptacle containing solid zinc oxide retained therein. 14. A method in accordance with claim 13 in which said solid zinc oxide is a bed of zinc oxide particles. 15. A method in accordance with claim 14 in which said zinc oxide particles are sintered zinc oxide. 16. A method for inhibiting the deposition of radioactive cobalt in a water-bearing vessel of a water-cooled nuclear reactor, said method comprising: 17. A method in accordance with claim 16 in which the zinc oxide content of said slurry is from about 0.1% to about 20% by weight. 18. A method in accordance with claim 16 in which the zinc oxide content of said slurry is from about 1% to about 5% by weight. 19. A method in accordance with claim 16 in which said zinc oxide in said slurry is fumed zinc oxide. 20. A method in accordance with claim 16 in which the rate of addition of said slurry to said water is selected to produce a zinc content in the water of said water-bearing vessel of about 1 to about 1,000 ppb. 21. A method in accordance with claim 16 in which the rate of addition of said slurry to said water is selected to produce a zinc content in the water of said water-bearing vessel of about 3 to about 100 ppb. |
claims | 1. A radiation window membrane for covering an opening in an X-ray device, through which opening X-rays are to pass, the membrane comprising:a window base layer, anda pinhole-blocking layer on a surface of said window base layer;wherein said pinhole-blocking layer comprises graphene. 2. A radiation window membrane according to claim 1, wherein said pinhole-blocking layer is electrically conductive. 3. A radiation window membrane according to claim 1, wherein said window base layer comprises at least one of: aluminium oxide, aluminium nitride, titanium oxide, silicon nitride. 4. A radiation window membrane according to claim 3, comprising a patterned layer on one side of said pinhole-blocking layer, wherein said patterned layer is one of the following: a patterned copper layer, a patterned nickel layer, a patterned iridium layer, a patterned ruthenium layer. 5. A radiation window membrane according to claim 4, wherein said pinhole-blocking layer is on one side of said patterned layer, and the radiation window membrane comprises a patterned substrate on another side of said patterned layer. 6. A radiation window membrane according to claim 3, comprising an etch stop layer on a different side of said window base layer than said pinhole-blocking layer. 7. A radiation window membrane according to claim 1, wherein the radiation window membrane comprises additionally a support layer, which is one of: a continuous polymer film, a support mesh made of polymer, a support mesh made of metal. 8. A method for manufacturing a radiation window membrane for covering an opening in an X-ray device, through which opening X-rays are to pass, the method comprising:attaching a pinhole-blocking layer to a window base layer;wherein said pinhole-blocking layer comprises graphene. 9. A method according to claim 8, comprising:using a thin film manufacturing technique to produce a graphene layer on an etchable support layer, wherein said graphene layer constitutes said pinhole-blocking layer,using a thin film manufacturing technique to produce a window base layer on said graphene layer, andetching through said etchable support layer to leave a patterned support layer on one side of said graphene layer. 10. A method according to claim 9, comprising:before producing the graphene layer, using a thin film manufacturing technique to produce said etchable support layer on an etchable substrate layer, andin said etching step, etching through both the etchable substrate layer and said etchable support layer. 11. A method according to claim 8, comprising:producing a first membrane, which comprises an exposed graphene layer,producing a second membrane, which comprises an exposed window base layer, andattaching said first membrane to said second membrane, so that said exposed graphene layer becomes attached to said exposed window base layer. 12. A method according to claim 11, wherein:producing said first membrane comprises using a thin film manufacturing technique to produce said graphene layer on a first support layer, producing a second support layer on a different surface of said graphene layer than said first support layer, and removing the first support layer to expose said graphene layer. 13. A method according to claim 11, wherein:producing said second membrane comprises using a thin film manufacturing technique to produce an etch stop layer on a substrate, and using a thin film manufacturing technique to produce said window base layer on said etch stop layer. 14. A method according to claim 13, wherein:producing said first membrane comprises using a thin film manufacturing technique to produce said graphene layer on a first support layer, producing a second support layer on a different surface of said graphene layer than said first support layer, and removing the first support layer to expose said graphene layer,after attaching said first membrane to said second membrane, the method comprises removing at least part of said second support layer and at least part of said substrate. |
|
abstract | A hierarchical representation encapsulates the detailed internal composition of a sub-circuit using the notion of a cell definition (a CellDef). The CellDef serves as a natural unit for operational reuse. If the computation required for the analysis or manipulation (e.g. parasitic extraction, RET, design rule confirmation (DRC), or OPC) based on a CellDef or one cell instance can be applied, with no or minimal additional effort, to all or a significant subset of other instances of the cell, very substantial reduction in computational effort may be realized. Furthermore, a hierarchical representation also allows for the partitioning of the overall analysis/manipulation task into a collection of subtasks, e.g. one per CellDef. Multiple jobs may then be distributed across a large number of computational nodes on a network for concurrent execution. While this may not reduce the aggregate computational time, a major reduction in the overall turnaround time (TAT) is in itself extremely beneficial. |
|
053751507 | summary | A portion of the disclosure of this patent document contains material which is subject to copyright protection. The copyright owner has no objection to the facsimile reproduction by anyone of the patent document or the patent disclosure, as it appears in the Patent and Trademark Office patent file or records, but otherwise reserves all copyright rights whatsoever. BACKGROUND OF THE INVENTION The present invention relates to apparatus and methods for monitoring and controlling the operation of commercial nuclear power plants. Conventionally, commercial nuclear power plants have a central control room containing equipment by which the operator collects, detects, reads, compares, copies, computes, complies, analyzes, confirms, monitors, and/or verifies many bits of information from multiple indicators ad alarms. Conventionally, the major operational systems in the control room have been installed and operate somewhat independently. These include the monitoring function, by which the components and the various processes in the plant are monitored; control, by which the components and the processes are intentionally altered or adjusted, and protection, by which a threat to the safety of the plant is identified and corrective measures immediately taken. The result of such conventional control room arrangement and functionality can sometimes be information overload or stimulus overload on the operator. That is, the amount of information and the variety and complexity of the equipment available to the operator for taking action based on such extensive information, can exceed the operator's cognitive limits, resulting in errors. The most famous example of the inability of operators to assimilate and act correctly based on the tremendous volume of information stimuli in the control room, particularly during unexpected or unusual plant transients, is the accident that occurred in 1978 at the Three Mile Island nuclear power plant. Since that event, the industry has focused considerable attention to increasing plant operability through improving control room operator performance. A key aspect to the improvement process is the use of human engineering design principles. Advances in computer technology since 1978 have enabled nuclear engineers and control room designers to display more information, in a greater variety of ways, but this can be counterproductive, because part of the problem is the overload of information. Improving "user friendliness" while maintaining the quantity and type of information at the operator's disposal has posed a formidable engineering challenge. SUMMARY OF THE INVENTION It is thus an object of the present invention to provide apparatus and method for nuclear power plant control and monitoring operations having the characteristics of concise information processing and display, reliable architecture and hardware, and easily maintainable components, while eliminating operator information overload. This objective should be accomplished while achieving enhanced reliability, ease of operation, and overall cost effectiveness of the control room complex. The solution to the problem is accomplished with the present invention by providing a number of features which are novel both individually and as integrated together in a control complex. The complex includes six major systems: (1) the control center panels, (2) the data processing system (DPS), (3) the discrete indication and alarm system (DIAS), (4) the component control system consisting of the engineered safeguard function component controls (ESFC) and the process component controls (PCC), (5) the plant protection system (PPS), and (6) the power control system (PCS). These six systems collect data from the plant, efficiently present the required information to the operator, perform all automatic functions and provide for direct manual control of the plant components. The control complex in accordance with the invention provides a top-down integrated information display and alarm approach that supports rapid assessment of high level critical plant safety and power production functions; provides guidance to the operator regarding the location of information to further diagnose high level assessments; and significantly reduces the number of display devices relative to conventional nuclear control complexes. The complex also significantly reduces the amount of data the operator must process at any one time; significantly reduces the operational impact of display equipment failures; provides fixed locations for important information; and eliminates display system equipment used only for off normal plant conditions. It is known that the nuclear stream supply system can be kept in a safe, stable state by maintaining a limited set of critical safety functions. The present invention extends the concept of the critical plant safety functions to include critical plant power production functions, in essence integrating the two functions so that the information presentation to the operator supports all high level critical plant functions necessary for power production as well as safety. The information display hierarchy in accordance with the invention includes a "big board" integrated process status overview screen (IPSO) at the apex, which provides a single dedicated location for rapid assessment of key information indicative of critical plant power production and safety functions. Further detail on the sources and trends of normal or abnormal parameter changes are provided by the DIAS. Both IPSO and the DIAS provide direct access and guidance to additional system and component status information contained on a hierarchy of CRT display pages which are driven by the DPS. The IPSO continually displays spatially dedicated information that provides the status of the plant's critical safety and power production functions. This information is presented using a small number of easily understood symbolic representations that are the results of highly processed data. This relieves the operator of the burden of correlating large quantities of individual parameter data, systems or component status, and alarms to ascertain the plant functional conditions. The IPSO presents the operator with high level effects of lower level component problems. The IPSO relies primarily on parameter trend direction, e.g., higher, lower, an alarm symbol color and shape, to convey key information. These are supplemented by values for selected parameters. The IPSO presents consolidated, simplified information to the operator in relatively small quantities of easily recognized and understood information. Furthermore, the IPSO compensates for the disadvantage inherent in recent industry trends towards presenting all information serially on CRTs, by enabling the operator to obtain an overview, or "feel" of the plant condition. Display of plant level overview on a large-format dedicated display addresses two additional operational concerns. First, operator tasks often require detailed diagnostics in very limited process areas. However, maintaining concurrent awareness of plant-wide performance is also necessary. Rather than relying on multiple operators in the control room to monitor respective indicators and the like on spatially separated panels, the IPSO can be viewed from anywhere in the control room and thus provides an operator a continuous indication of plant performance regardless of the detailed nature of the task that may be requiring the majority of his attention. In the preferred implementation, IPSO supports the assessment of the power ad safety critical functions by providing for each function, key process parameters that indicate the functional status. For each function, key success paths are selected with the status of the success path displayed. The IPSO clearly relates functions to physical things in the plant. The critical functions are applied to power production, normal post trip actions, and optimal functional recovery procedures. The second level in the display information hierarchy in accordance with the present invention is the presentation of plant alarms from the DIAS. A limited number of fixed, discrete tiles are used with three levels of alarm priorities. Dynamic alarm processing uses information about the plant state (e.g., reactor power, reactor trip, refueling, shut-down, etc.) and information about system and equipment status to eliminate unnecessary and redundant alarms that would otherwise contribute to operator information overload. The alarm system provides a supplementary level of easily understood cueing into further information in the discrete indicators, CRTs and controls. Alarms are based on validated data, so that the alarms indentify real plant process problems, not instrumentation and control system failures. The alarm features include providing a detailed message through a window to the operator upon the acknowledgement of an alarm and the ability to group the alarms without losing the individual messages. The tiles can dynamically display different priorities to the operator. The acknowledgement sequence ensures that all alarms are acknowledges while at the same time reducing the operator task loading by providing momentary tones, then continuous alarm, followed by reminder tones to ensure that the alarms are not forgotten. The operator has the ability to stop temporarily alarm flashing to avoid visual overload, and resume the flashing to ensure that the alarm will eventually be acknowledged. The discrete indicators in the DIAS provide the third level of display in the hierarchy of the present invention. The flat panel displays compress many signal sources into a limited set of read-outs for frequently monitored key plant data. Signal validation and automatic selection of sensors with the most accurate signal ranges are also employed to reduce the number of control panel indicators. Information read-outs are by touch-screen to enhance operator interaction and include numeric parameter values, a bar form of analog display, and a plot trend. Various multi-range indicators are available on one display with automatic sensor selection and range display. The automatic calculation of a valid process representation parameter value, with the availability of individual sensor readings at the same display, avoids the need for separate backup displays, or different displays or normal operation versus accident or post-accident operation. Moreover, in another preferred feature o the invention, the parameter verification automatically distinguishes failed or multiple failed sensors, while allowing continued operation and accident mitigation information to the operator even if the CRT display is not available. Furthermore, the normal display information can be correlated to a qualified sensor, such as that used for post-accident monitoring purposes. At the information display level associated with control of specific components, dynamic "soft" controllers are provided with component status and control signal information necessary for operator control of these components. For the ESFC system, this information includes status lamp, on-off controls, modulation controls, open-closed controls, and logic controls. For the PCCS, the information includes confirm load, set points, operating range, process values, and control signal outputs. In the fourth level of the information hierarchy, dynamic CRT display pages are complementary to all levels of spatially dedicated control and information and can be accessed from ay CRT location in the control room, technical support center, or emergency operations facility. These displays are grouped into a three level hierarchy that includes general monitoring (level 1), plant component and systems control (level 2), and component/process diagnostics (level 3). Display implementation is driven by the DPS and duplicates and verifies all discrete alarm and indicator processing performed in the DIAS. In the preferred implementation of the invention, the indicator, alarm, and control functions for a given major functional system of the plant are grouped together in a single, modularized panel. The panel can be made with cutouts that are spatially dedicated to each of the displays for the indicators, alarms, controls, and CRT, independent of the major plant functional system. This permits delivery, installation, and preliminary testing of the panels before finalization of the plant specific logic and algorithms, which can be software modified late in the plant construction schedule. This modularization is achievable because the space required on the panel is essentially independent of the major plant functional system to which the panel is dedicated. Both the alarms and indicators can be easily modified in software. The number of indicators and alarm tiles that can be displayed to the operator are not significantly limited by the available area of the panel, so that stanardization of panel size and cutout locations for the display windows is possible. |
description | 1. Field of the Invention The present invention relates to a lithographic projection apparatus and device manufacturing method. 2. Description of the Related Art The term “patterning device” as here employed should be broadly interpreted as referring to device that can be used to endow an incoming radiation beam with a patterned cross-section, corresponding to a pattern that is to be created in a target portion of the substrate. The term “light valve” can also be used in this context. Generally, the pattern will correspond to a particular functional layer in a device being created in the target portion, such as an integrated circuit or other device (see below). An example of such a patterning device is a mask. The concept of a mask is well known in lithography, and it includes mask types such as binary, alternating phase-shift, and attenuated phase-shift, as well as various hybrid mask types. Placement of such a mask in the radiation beam causes selective transmission (in the case of a transmissive mask) or reflection (in the case of a reflective mask) of the radiation impinging on the mask, according to the pattern on the mask. In the case of a mask, the support will generally be a mask table, which ensures that the mask can be held at a desired position in the incoming radiation beam, and that it can be moved relative to the beam if so desired. Another example of a patterning device is a programmable minor array. One example of such an array is a matrix-addressable surface having a viscoelastic control layer and a reflective surface. The basic principle behind such an apparatus is that, for example, addressed areas of the reflective surface reflect incident light as diffracted light, whereas unaddressed areas reflect incident light as undiffracted light. Using an appropriate filter, the undiffracted light can be filtered out of the reflected beam, leaving only the diffracted light behind. In this manner, the beam becomes patterned according to the addressing pattern of the matrix-addressable surface. An alterative embodiment of a programmable mirror array employs a matrix arrangement of tiny mirrors, each of which can be individually tilted about an axis by applying a suitable localized electric field, or by employing piezoelectric actuators. Once again, the mirrors are matrix-addressable, such that addressed mirrors will reflect an incoming radiation beam in a different direction to unaddressed mirrors. In this manner, the reflected beam is patterned according to the addressing pattern of the matrix-addressable mirrors. The required matrix addressing can be performed using suitable electronics. In both of the situations described hereabove, the patterning device can comprise one or more programmable mirror arrays. More information on mirror arrays as here referred to can be seen, for example, from U.S. Pat. No. 5,296,891 and 5,523,193, and PCT publications WO 98/38597 and WO 98/33096. In the case of a programmable mirror array, the support may be embodied as a frame or table, for example, which may be fixed or movable as required Another example of a patterning device is a programmable LCD array. An example of such a construction is given in U.S. Pat. No. 5,229,872. As above, the support structure in this case may be embodied as a frame or table, for example, which may be fixed or movable as required. For purposes of simplicity, the rest of this text may, at certain locations, specifically direct itself to examples involving a mask and mask table. However, the general principles discussed in such instances should be seen in the broader context of the patterning device as hereabove set forth. Lithographic projection apparatus can be used, for example, in the manufacture of integrated circuits (IC's). In such a case, the patterning device may generate a circuit pattern corresponding to an individual layer of the IC, and this pattern can be imaged onto a target portion (e.g. comprising one or more dies) on a substrate (silicon wafer) that has been coated with a layer of radiation-sensitive material (resist). In general, a single wafer will contain a whole network of adjacent target portions that are successively irradiated via the projection system, one at a time. In current apparatus, employing patterning by a mask on a mask table, a distinction can be made between two different types of machine. In one type of lithographic projection apparatus, each target portion is irradiated by exposing the entire mask pattern onto the target portion at once. Such an apparatus is commonly referred to as a wafer stepper. In an alternative apparatus, commonly referred to as a step-and-scan apparatus, each target portion is irradiated by progressively scanning the mask pattern under the projection beam in a given reference direction (the “scanning” direction) while synchronously scanning the substrate table parallel or anti-parallel to this direction. Since, in general, the projection system will have a magnification factor M (generally <1), the speed V at which the substrate table is scanned will be a factor M times that at which the mask table is scanned. More information with regard to lithographic devices as here described can be seen, for example, from U.S. Pat. No. 6,046,792. In a known manufacturing process using a lithographic projection apparatus, a pattern (e.g. in a mask) is imaged onto a substrate that is at least partially covered by a layer of radiation-sensitive material (resist). Prior to this imaging, the substrate may undergo various procedures, such as priming, resist coating and a soft bake. After exposure, the substrate may be subjected to other procedures, such as a post-exposure bake (PEB), development, a hard bake and measurement/inspection of the imaged features. This array of procedures is used as a basis to pattern an individual layer of a device, e.g. an IC. Such a patterned layer may then undergo various processes such as etching, ion-implantation (doping), metallization, oxidation, chemo-mechanical polishing, etc., all intended to finish off an individual layer. If several layers are required, then the whole procedure, or a variant thereof, will have to be repeated for each new layer. It is important to ensure that the overlay juxtaposition) of the various stacked layers is as accurate as possible. For this purpose, a small reference mark is provided at one or more positions on the wafer, thus defining the origin of a coordinate system on the wafer. Using optical and electronic devices in combination with the substrate holder positioning device (referred to hereinafter as “alignment system”), this mark can then be relocated each time a new layer has to be juxtaposed on an existing layer, and can be used as an alignment reference. Eventually, an array of devices will be present on the substrate (wafer). These devices are then separated from one another by a technique such as dicing or sawing, whence the individual devices can be mounted on a carrier, connected to pins, etc. Further information regarding such processes can be obtained, for example, from the book “Microchip Fabrication: A Practical Guide to Semiconductor Processing”, Third Edition, by Peter van Zant, McGraw Hill Publishing Co., 1997, ISBN 0-07-067250-4. For the sake of simplicity, the projection system may hereinafter be referred to as the “lens.” However, this term should be broadly interpreted as encompassing various types of projection system, including refractive optics, reflective optics, and caradioptric systems, for example. The radiation system may also include components operating according to any of these design types for directing, shaping or controlling the beam of radiation, and such components may also be referred to below, collectively or singularly, as a “lens”. Further, the lithographic apparatus may be of a type having two or more substrate tables (and/or two or more mask tables). In such “multiple stage” devices the additional tables may be used in parallel or preparatory steps may be carried out on one or more tables while one or more other tables are being used for exposures. Dual stage lithographic apparatus are described, for example, in U.S. Pat. No. 5,969,441 and WO 98/40791. In order to accurately expose the substrate, the projection system must be mounted on the apparatus in such a manner that any vibrations or displacements within the apparatus do not affect the position of the elements within the projection system. This is critical since the projection system contains many optical elements connected to a projection frame which must be accurately positioned relative to one another and relative to other components in the apparatus. Displacements and vibrations of the projection frame may cause so-called rigid body errors in which the projection frame as a whole is displaced and the optical elements within it are displaced relative to other components in the apparatus. In addition the vibrations and distortions of the projection frame may cause flexure and/or strains in the projection frame which reduce the accuracy of the relative positions of the elements mounted on the projection frame. Both sources of positional inaccuracy of the optical elements (relative to the remainder of the apparatus and relative to the other elements) reduce the accuracy of the projection of the image onto the substrate. It has therefore been previously known to rigidly mount the projection frame onto a reference frame which is compliantly mounted on the base of the apparatus. Such a reference frame is used to support components of the apparatus, such as position sensors, that are sensitive to vibrations. By compliantly mounting, or “soft-mounting”, the reference frame on the base, vibrations and displacements acting on the base (which may be caused by, for example, reaction forces from actuators positioning other components of the apparatus) will be low-pass filtered by the mounts and the disturbance on the reference frame, and hence the projection frame and the optical elements, is reduced. However, as the critical dimension of devices produced using the lithography process decreases, the required accuracy of the apparatus increases. Therefore, the continuing increase in demand for smaller device dimensions requires that the tolerable position error within the projection optics be reduced. In order to allow for this reduction, the soft-mount for the reference frame must be improved to reduce the vibration transmitted from the base of the apparatus to the reference frame. In conventional lithographic apparatus the soft-mounts may be so-called “air-mounts” in which the reference frame is supported on an air cushion of some restricted volume. In order to improve such an air-mount to meet the requirements for the maximum vibration being passed to the projection optics, the air-mounts supporting the reference frame must be increased in size. However the space available around the lithographic projection apparatus is limited and, therefore, when the required performance for the mounts (driven by the dimension of the devices to be formed on the substrate) reaches a given value, the conventional arrangements supporting the projection frame are no longer appropriate. Furthermore, even where the reference frame is sufficiently isolated from the base, additional vibrations may be introduced to the reference frame by, for example, water cooling tubes. It is an aspect of the present invention to provide a lithographic apparatus in which the projection frame is supported in such a manner that the vibration and positional errors transmitted to it are reduced or eliminated. This and other aspects are achieved according to the present invention in a lithographic apparatus including a radiation system constructed and arranged to provide a beam of radiation; a support constructed and arranged to support a patterning device, the patterning device constructed and arranged to pattern the projection beam according to a desired pattern; a substrate table constructed and arranged to hold a substrate; a projection system constructed and arranged to project the patterned beam onto a target portion of the substrate; a base to which the support and the substrate table are mounted; and a reference frame, compliantly mounted to the base, wherein the projection system comprises at least one optical element mounted on a projection frame that is compliantly mounted to the reference frame. The present invention limits vibrations transmitted to the projection system by using mounts for the reference frame which are not larger than conventional mounts. Other elements of the lithographic apparatus, which do not need this level of positional accuracy, may be mounted to the reference frame in the conventional manner. Preferably, the eigenfrequency of the mounting of the projection frame on the reference frame is between about 10 and 30 Hz and the eigenfrequency of the mounting of the reference frame on the base is about 0.5 Hz. This results in very low levels of vibration being transmitted from the base to the projection system, but the size of the mounts for the reference frame will still be within practical limits. The motion of the projection frame relative to the reference frame may be damped. This reduces the displacement of the projection frame relative to the reference frame at the low frequencies around the resonant frequency of the mounting. In a preferred embodiment of the present invention, the projection frame is mounted on the reference frame using compliant mounts that are attached to the reference frame on the nodal axes of the dominant mode of bending vibration or torsional vibration of the reference frame. As such bending or torsional vibration of the reference frame produces the greatest distortion of the reference frame but, by mounting it in this fashion, very little of this distortion will be transmitted to the projection frame. In one embodiment the compliant mounts used to mount the projection frame onto the reference frame are at least one of an air-mount, a spring and a magnetic support. These mounts have responses that are predictable. According to an alternative embodiment, the compliant mounts are formed from a T-shaped member with one of the projection frame and reference frame attached to both ends of the cross-piece and the other attached to the end of the leg. This compliant mount requires very little maintenance, can easily be replaced and may be simply tuned by adjusting the lengths and widths of the cross-piece and leg members. According to a further preferred embodiment, the T-shaped member has an internal first eigenfrequency that is greater than 1000 Hz. In this way vibrational problems related to the suspension of the projection system can be avoided. According to a further aspect of the invention there is provided a device manufacturing method including projecting a patterned beam of radiation onto a target portion of a layer of radiation-sensitive material at least partially covering a substrate using a projection system; supporting a reference frame, a support constructed and arranged to support a patterning device and a substrate table constructed and arranged to hold the substrate on a base, wherein the reference frame is compliantly mounted to the base and the projection system is mounted to the reference frame; and compliantly mounting the projection system to the reference frame while projecting the patterned beam of radiation onto the target portion. Although specific reference may be made in this text to the use of the apparatus according to the invention in the manufacture of ICs, it should be explicitly understood that such an apparatus has many other possible applications. For example, it may be employed in the manufacture of integrated optical systems, guidance and detection patterns for magnetic domain memories, liquid-crystal display panels, thin-film magnetic heads, etc. One of ordinary skill in the art will appreciate that, in the context of such alternative applications, any use of the terms “reticle”, “wafer” or “die” in this text should be considered as being replaced by the more general terms “mask”, “substrate” and “target portion”, respectively. In the present document, the terms “radiation” and “beam” are used to encompass all types of electromagnetic radiation, including ultraviolet radiation (e.g. with a wavelength of 365, 248, 193, 157 or 126 nm) and EUV (extreme ultra-violet radiation, e.g. having a wavelength in the range 5–20 nm), as well as particle beams, such as ion beams or electron beams. In the Figures, corresponding reference symbols indicate corresponding parts. FIG. 1 schematically depicts a lithographic projection apparatus 1 according to an embodiment of the invention. The apparatus 1 includes a base BP; a radiation system Ex, IL constructed and arranged to supply a beam PB of radiation (e.g. EUV radiation), which in this particular case also comprises a radiation source LA; a first object (mask) table MT provided with a mask holder constructed and arranged to hold a mask MA (e.g. a reticle), and connected to a first positioning device PM that accurately positions the mask with respect to a projection system or lens PL; a second object (substrate) table WT provided with a substrate holder constructed and arranged to hold a substrate W (e.g. a resist-coated silicon wafer), and connected to a second positioning device PW that accurately positions the substrate with respect to the projection system PL. The projection system or lens PL (e.g. a mirror group) is constructed and arranged to image an irradiated portion of the mask MA onto a target portion C (e.g. comprising one or more dies) of the substrate W. As here depicted, the apparatus is of a reflective type (i.e. has a reflective mask). However, in general, it may also be of a transmissive type, for example with a transmissive mask. Alternatively, the apparatus may employ another kind of patterning device, such as a programmable mirror array of a type as referred to above. The source LA (e.g. a discharge or laser-produced plasma source) produces radiation. This radiation is fed into an illumination system (illuminator) IL, either directly or after having traversed a conditioning device, such as a beam expander Ex, for example. The illuminator IL may comprise an adjusting device AM that sets the outer and/or inner radial extent (commonly referred to as σ-outer and σ-inner, respectively) of the intensity distribution in the beam. In addition, it will generally comprise various other components, such as an integrator IN and a condenser CO. In this way, the beam PB impinging on the mask MA has a desired uniformity and intensity distribution in its cross-section. It should be noted with regard to FIG. 1 that the source LA may be within the housing of the lithographic projection apparatus, as is often the case when the source LA is a mercury lamp, for example, but that it may also be remote from the lithographic projection apparatus, the radiation beam which it produces being led into the apparatus (e.g. with the aid of suitable directing mirrors). This latter scenario is often the case when the source LA is an excimer laser. The present invention encompasses both of these scenarios. The beam PB subsequently intercepts the mask MA, which is held on a mask table MT. Having traversed the mask MA, the beam PB passes through the lens PL, which focuses the beam PB onto a target portion C of the substrate W. With the aid of the second positioning device PW and interferometer IF, the substrate table WT can be moved accurately, e.g. so as to position different target portions C in the path of the beam PB. Similarly, the first positioning device PM can be used to accurately position the mask MA with respect to the path of the beam PB, e.g. after mechanical retrieval of the mask MA from a mask library, or during a scan. In general, movement of the object tables MT, WT will be realized with the aid of a long-stroke module (coarse positioning) and a short-stroke module (fine positioning), which are not explicitly depicted in FIG. 1. However, in the case of a wafer stepper (as opposed to a step and scan apparatus) the mask table MT may just be connected to a short stroke actuator, or may be fixed. The mask MA and the substrate W may be aligned using mask alignment marks M1, M2 and substrate alignment marks P1, P2. The depicted apparatus can be used in two different modes: 1. In step mode, the mask table MT is kept essentially stationary, and an entire mask image is projected at once, i.e. a single “flash,” onto a target portion C. The substrate table WT is then shifted in the X and/or Y directions so that a different target portion C can be irradiated by the beam PB; 2. In scan mode, essentially the same scenario applies, except that a given target portion C is not exposed in a single “flash.” Instead, the mask table MT is movable in a given direction (the so-called “scan direction”, e.g., the Y direction) with a speed v, so that the beam PB is caused to scan over a mask image. Concurrently, the substrate table WT is simultaneously moved in the same or opposite direction at a speed V=Mv, in which M is the magnification of the lens Pl. (typically. M=¼ or ⅕). In this manner, a relatively large target portion C can be exposed, without having to compromise on resolution. FIG. 2 shows the projection system PL mounted to an isolated reference frame 5. The reference frame 5 is, in turn, mounted on the base BP of the apparatus. Compliant mounts 6 support the reference frame 5. The mounts 6 have low-pass characteristics and may have an eigenfrequency of between about 0.1 Hz and 10 Hz and preferably have an eigenfrequency of approximately 0.5 Hz. The mounts 6 therefore attenuate the transmission of any vibrations or displacements from the base BP to the reference frame 5. The attenuation of vibrations is necessary as the vibrations in the base BP will be larger than the maximum allowed in the projection system PL. The vibrations in the base frame may be caused by, for example, reaction forces from the actuators used to position the elements of the apparatus. However, despite the use of the mounts 6, the amplitude of the vibrations in the reference frame 5 may be larger than the maximum allowable in the projection system PL. Therefore the projection system PL is mounted to the reference frame 5 by compliant mounts 7. The mounts 7 present low-pass characteristics, having cut-off or eigenfrequencies of, for example, between about 10 and about 30 Hz. The soft-mounting of the projection system PL on the reference frame 5 therefore further filters the vibration and motion of the base BP. FIG. 3 shows a projection system PL that may be used with the present invention. It comprises a projection frame PF, to which are mounted a plurality of optical elements M1, M2, M3, M4, M5, M6. The projection system of the present invention may further comprise a second projection frame (not shown in FIG. 2) to which sensors may be attached to monitor the position of the optical elements M1, M2, M3, M4, M5, M6. One or, preferably, both of the projection frames may be compliantly mounted on the reference frame. If only one projection frame is to be compliantly mounted on the reference frame, it may be desirable for this to be the second projection frame which then ensures a fully stable frame of reference to monitor the position of the optical elements M1, M2, M3, M4, M5, M6 within the projection system PL. Additionally compliantly mounting the first projection frame on the reference frame prevents force feedback through the reference frame. The reference frame mounts 6 and the projection system mounts 7 may be any form of so-called soft-mounts. That is to say, the mounts 6, 7 may be any kind of mount with low stiffness. For instance, the mounts may be very soft springs, air-mounts (in which one part of the mount is supported relative to the other by a cushion of a volume of gas) or may be magnetic mounts (in which one part of the mount is supported relative to the other by a magnetic force). The mounts 6, 7, especially the projection system mounts 7, may, alternatively, be solid mounts. FIG. 6 shows a design of such a solid mount 10. The mount 10 includes a first elongate member 11 which is attached, by attachment points 13, at either end, to one of the reference frame 5 and the projection frame PF. A second elongate member 12 is connected to the first elongate member 11 at one end 12a and has an attachment point 14 at the other end 12b. The attachment point 14 is used to connect the mount 10 to the other of the reference frame 5 and the projection frame PF. The second elongate member 12 is connected to the first elongate member 11 between the two attachment points 13 on the first elongate member. Consequently, bending and flexure of the first and second elongate members 11, 12 provide relative movement between the attachment points 13 on the first elongate member and the attachment point 14 on the second elongate member. The frequency response of the mount 10 can be selected by using materials with appropriate stiffness and altered by adjusting the width and length of the first and second elongate members 11, 12. The mount may be formed from, for example, high tensile steel or Invar™ (a cobalt containing steel that has a relatively low coefficient of thermal expansion). It is preferred that the T-shaped mount 10 has an internal first eigenfrequency that is greater than 1000 Hz. In particular, it is preferred that the internal first eigenfrequency of the mount 10 is much larger than 1000 Hz. As shown in FIG. 6, there is a space 11a below the first elongate member 11 and between the two attachment points 13 on the first elongate member 11. A block (not shown) may be provided in this space 11a to prevent large movements of the first elongate member 11. Compliantly mounting the projection system PL onto the reference frame 5 may, however, result in displacement errors of the projection system PL relative to the reference frame 5. This will be especially true for vibrations that are close to the eigenfrequency of the mount. Therefore it may be necessary to provide damping of the motion of the projection system PL relative to the reference frame 5. The damping may be provided within the mounts themselves or may be applied at a different part of the projection frame PF. The damping applied may be passive or may be active using, for example, piezoelectric actuators or Lorentz-force actuators. The choice of location of the mounts on the reference frame may also aid to reduce the amount of vibration and position error that is transmitted to the projection system PL. FIG. 4 shows a possible mounting location 15 for the projection system on the reference frame 5. The reference frame 5 may have torsional modes of vibration in which two axes 16, 17 in the plane of the frame are nodal. That is to say, when the reference frame 5 vibrates purely in this mode, the points on these nodal axes 16, 17 remain substantially motionless. Therefore by attaching the mounts at these locations 15, substantially no vibrations from this mode of vibration are transmitted to the projection system PL. FIG. 5 shows an alternative arrangement in which the mounts are attached at locations 18 on the nodal axes for the dominant bending mode of vibration for the reference frame 5. Again, when the reference frame 5 vibrates predominantly in this mode, the nodal axes 19, 20 are substantially motionless and therefore vibrations in this mode are not transmitted to the projection system PL. The projection system may be mounted according to the configuration shown in FIG. 3 or FIG. 4, depending on which mode of vibration of the reference frame 5 (torsional or bending) is the most detrimental to the projection system. In addition to the projection system, other components which are sensitive to vibration and displacement errors may be mounted on the reference frame. In particular, measurement sensors, for determining the position of, for example, the substrate table or the mask holder may be mounted on the reference frame. These components may also be mounted on the nodal axes of the reference frame. While specific embodiments of the invention have been described above, it will be appreciated that the invention may be practiced otherwise than as described. The description is not intended to limit the invention. |
|
description | This invention was made with government support under Contract No. DE-AC07-05ID-14517 awarded by the United States Department of Energy. The government has certain rights in the invention. This invention relates to nuclear reactor systems in general and, more specifically, to systems and methods for providing large neutron fluxes and high fast-to-thermal neutron ratios. It is often desirable in nuclear reactor systems to produce large or substantial neutron fluxes and high fast-to-thermal neutron ratios, either in order to test and analyze the performance of new reactor fuels and materials in such environments or to otherwise operate nuclear reactor systems under “fast-flux” radiation conditions. One method for testing materials under fast-flux radiation conditions is to place the materials in a reactor and expose them to fast neutrons. More particularly, it is often desirable to conduct such testing under conditions of high fast-to-thermal neutron ratios (e.g., ratios of about 15 or greater) as well as at large or substantial neutron flux intensities (e.g., 1×1015 n/cm2·s or greater). Unfortunately, these neutron energy and flux conditions place conflicting technical requirements on the configuration and operational parameters of the reactor system. For example, the requirement for a high fast-to-thermal neutron ratio (e.g., in excess of 15) typically requires the absence of significant neutron thermalizer between the neutron source and the target material. In addition, neutron “filters” may need to be provided to remove or filter low energy or thermal neutrons from the neutron flux. The requirement for a high neutron flux intensity (e.g., in excess of 1015 n/cm2·s) typically results in substantial heating of the target material being studied, and may involve the use of booster fuels that further add to the heat load. In addition, such high flux intensities will result in additional heating of the neutron filter, e.g., from the (n,γ) absorption reactions, which additional heat must also be somehow removed from the apparatus. While numerous types of cooling systems are known and may be used for this purpose, they too, present conflicting design requirements. For example, while gas cooling systems are known and may be used, they must be operated at substantial pressures and flow rates in order to remove the excessive heat generated as a result of the high neutron flux intensities. Past experience has indicated that gas cooling systems are expensive and difficult to operate in such regimes. While molten metal and molten salt cooling systems are also known and may be used, they are not without their problems. For example, the use of liquid metals and salts may present safety concerns if they are reactive with the primary coolant being used in the test reactor. Moreover, such molten coolants (e.g., either metals or salts), must be maintained in the molten state in order to avoid structural damage to the system. Water cooled systems are also known and could be used. However, water is a highly effective neutron thermalizer, and thus serves to lower the fast-to-thermal neutron flux ratio. Consequently, the task of designing a reactor system suitable for exposing materials to high fast-to-thermal neutron flux ratios and at high flux intensities is by no means trivial and presents a number of conflicting technical and economic requirements that must be resolved in order to arrive at a successful system. One embodiment of a conduction cooled neutron absorber according to the teachings provided herein may include a metal matrix composite that comprises a metal having a thermal neutron cross-section of at least about 50 barns and a metal having a thermal conductivity of at least about 1 W/cm·K. Also disclosed is a method for producing a conduction cooled neutron absorber that comprises providing a powder mixture comprising a first powder component having a high thermal neutron cross-section and a second powder component having a high thermal conductivity; and consolidating the powder mixture to form the conduction cooled neutron absorber. One embodiment of an apparatus for providing a neutron flux having a high fast-to-thermal neutron ratio according to the teachings provided herein may include a source of neutrons that produces fast neutrons and thermal neutrons. A neutron absorber positioned adjacent the neutron source absorbs at least some of the thermal neutrons so that a region adjacent the neutron absorber has a fast-to-thermal neutron ratio of at least about 15. A coolant in thermal contact with the neutron absorber removes heat from the neutron absorber. A method for absorbing neutrons according to the teachings provided herein may include positioning a neutron absorber in a neutron field, the neutron absorber comprising a first material having a high thermal neutron cross-section dispersed within a second material having a high thermal conductivity; allowing neutrons from the neutron field to penetrate the neutron absorber, the neutron absorber absorbing at least some of the neutrons; and contacting a portion of the neutron absorber with a coolant, the coolant removing heat from the neutron absorber. One embodiment of a conduction cooled neutron absorber 10 is best seen in FIGS. 1 and 2 and may comprise a portion of a test cell 12 that is configured to hold a quantity of target or sample material 14 (FIG. 2) that is to be exposed to neutron radiation. More specifically, the test cell 12 is configured to be received by a test reactor (not shown) that produces or generates neutrons (shown schematically in FIGS. 1 and 2 as “n”) that are used to irradiate the target material 14. In the embodiment illustrated in FIGS. 1 and 2, the conduction cooled neutron absorber 10 takes the form of a generally elongate, cylindrical structure 16 having a plurality of bores 18 provided therein that is sized to receive the target or sample material 14. The conduction cooled neutron absorber 10 may also be provided with one or more cooling channels 20 therein through which are caused to flow a coolant 22 (FIG. 2) to remove heat generated as a result of neutron exposure. Alternatively, the conduction cooled neutron absorber 10 may comprise other shapes and configurations, as will be described in further detail below. In the various embodiments shown and described herein, the conduction cooled neutron absorber 10 is formed from a composite material that comprises various constituents that provide the conduction cooled neutron absorber 10 with certain advantageous properties and characteristics, some of which are described herein and others of which will become apparent to persons having ordinary skill in the art after having become familiar with the teachings provided herein. In one embodiment, the composite material comprises a first material or constituent having a large thermal neutron cross-section that is dispersed within a second material or constituent having a high thermal conductivity. Examples of the first material include any of a wide range of hafnium intermetallic phases, such as hafnium aluminide (Al3Hf). An example of the second material includes aluminum. The resulting composite material (e.g., containing hafnium aluminide intermetallic phase in aluminum) allows the conduction cooled neutron absorber 10 to provide high fast-to-thermal neutron ratios (e.g., as a result of the effective removal or filtering of thermal neutrons by the hafnium) while simultaneously allowing high neutron fluxes to be achieved without overheating (e.g., by the rapid and efficient transfer of heat to the coolant 22 by the aluminum). Referring now primarily to FIG. 6, a conduction cooled neutron absorber 10 (FIGS. 1 and 2) comprising the aluminum/hafnium composite material described herein may be produced or fabricated by mixing together hafnium aluminide powder 24 with aluminum powder 26 to form a powder mixture 28. The resulting powder mixture 28 may then be consolidated (e.g., by pressing or by extruding) in order to form a neutron absorber 10 having the desired configuration. Alternatively, other fabrication methods may be used, as will be described in further detail herein. As already mentioned, the conduction cooled neutron absorber 10 may comprise a portion of a test cell 12 that is configured to be placed within a reactor (not shown), so that the target or sample material 14 contained within the conduction cooled neutron absorber 10 will be exposed to a neutron field produced or generated by the reactor. Generally speaking, the neutron field produced by the reactor will comprise neutrons “n” (FIGS. 1 and 2) having a wide range of energies. Neutrons having energies below about 0.68 electron volts (eV) are referred to herein as “thermal” neutrons, whereas neutrons having energies greater than about 0.1 million electron volts (MeV) are referred to herein as “fast” neutrons. In the example applications shown and described herein, it is desired to test the target material 14 under “fast flux” conditions, i.e., under comparatively high neutron flux intensities and wherein the ratio of fast-to-thermal neutrons is high. More specifically, it is generally desired to expose the material 14 contained within the bores 18 of conduction cooled neutron absorber 10 to neutron flux intensities of at least about 1×1015 neutrons per square centimeter-second (n/cm2·s), and wherein the fast-to-thermal neutron ratio is at least about 15. That is, the neutron flux should comprise about 15 times as many fast neutrons as thermal neutrons. Indeed, and as will be described in further detail herein, computer modeling of various example configurations of the conduction cooled neutron absorber 10 indicates that high neutron flux intensities having fast-to-thermal neutron ratios as high as 50 can be readily achieved. The target material 14 may be irradiated by such “fast flux” neutrons by placing the sample material 14 within the bore(s) 18 provided within the conduction cooled neutron absorber 10. The resulting test cell assembly 12 may then be positioned within the reactor (not shown) and exposed to the neutron field. Once inserted into the reactor, neutrons “n” travel through the conduction cooled neutron absorber 10 before reaching the target material 14 contained therein. The hafnium provided within the conduction cooled neutron absorber 10 absorbs thermal neutrons, thereby increasing the fast-to-thermal ratio of neutrons “n” that ultimately reach the sample material 14. Heat resulting from the irradiation of the sample material 14, as well as heat generated by the neutron absorption reactions (e.g., (n,γ) reactions) occurring in the conduction cooled neutron absorber 10, is conducted primarily by the aluminum in the neutron absorber 10 to the cooling channels 20, whereupon it is transferred to the coolant 22. A significant advantage of the present invention is that it may be utilized to produce (e.g., by thermal neutron filtering) neutron radiation having high fast-to-thermal neutron ratios. The high fast-to-thermal neutron ratios achievable with the present invention are due primarily to the composition of the conduction cooled neutron absorber 10. More specifically, the hafnium in the hafnium aluminide intermetallic portion of the conduction cooled neutron absorber 10 has a large thermal neutron cross-section and is highly effective in absorbing significant quantities of low energy or thermal neutrons in the neutron flux generated by the reactor. Consequently, any of a wide range of fast-to-thermal neutron ratios may be achieved by designing the conduction cooled neutron absorber 10 so that an appropriate amount of hafnium (e.g., in the form of hafnium aluminide) is positioned between the neutron source and the target material 14. As already mentioned, neutron fields having high flux intensities typically involve the generation of significant amounts of heat. For example, significant quantities of heat may be generated within the target material 14 being tested, particularly where the target material 14 comprises a fuel. In addition, the hafnium contained within the conduction cooled neutron absorber 10 may also produce significant quantities of heat as a result of the capture of thermal neutrons. The heat produced by these heat sources is rapidly and efficiently conducted to the coolant 22 by the aluminum component of the conduction cooled neutron absorber 10. Consequently, the temperature of the sample material 14, as well as the temperature of the conduction cooled neutron absorber 10 itself, may be more easily maintained within desired ranges or below a predetermined maximum temperature. Another advantage of the present invention is that the superior thermal conductivity of the material forming the conduction cooled neutron absorber 10 allows the coolant channels 20 to be located farther away from the sample target material 14 than would otherwise be the case, i.e., compared to materials having lower thermal conductivities. Consequently, the present invention allows for water to be used as a coolant, even though water is not normally used as a coolant in such applications because it is a highly effective neutron thermalizer. That is, even though the coolant 22 may thermalize a significant number of neutrons “n,” the overall adverse effect on the fast-to-thermal neutron ratio is reduced because the coolant 22 may be located farther away from the target material 14 being tested. The increased distance between the coolant 22 and the target material 14 allows the hafnium contained within the conduction cooled neutron absorber 10 to absorb or filter more of the thermal neutrons produced by the coolant 22. In addition, the high thermal conductivity of the conduction cooled neutron absorber 10 will allow lesser quantities of water coolant to be used, thereby further enhancing the ability to produce neutron fluxes having high fast-to-thermal neutron ratios. Still yet other advantages are associated with the present invention. For example, the powder metallurgy process for dispersing the hafnium intermetallic material (e.g., hafnium aluminide) within the aluminum matrix allows hafnium to be readily incorporated in aluminum, even though hafnium itself is not readily soluble in aluminum. The hafnium/aluminum composite material is also stable, both during fabrication and irradiation, and is resistant to corrosion. Moreover, the powder metallurgy techniques described herein may also allow the material to be formed into any of a wide variety of shapes and configurations. Having briefly described one embodiment of a conduction cooled neutron absorber 10 according to the present invention, as well as some of its more significant features and advantages, various exemplary embodiments of conduction cooled neutron absorbers, methods for producing “fast-flux” neutron fields, and methods for fabricating neutron absorbers will now be described in detail. However, before proceeding with the detailed description, it should be noted that the various embodiments of the conduction cooled neutron absorbers are shown and described herein as they could be used in test cell assemblies for exposing sample or target material 14 to fast-flux radiation fields produced by a test reactor. However, the present invention is not limited to such applications, and could instead be used in any of a wide range of applications wherein it is desired to filter thermal neutrons (e.g., to produce neutron fluxes having high fast-to-thermal neutron ratios) and/or wherein it is desired to provide high thermal conductivities. Consequently, the present invention should not be regarded as limited to the particular configurations and environments shown and described herein. Referring back now to FIGS. 1 and 2, a first embodiment of a conduction cooled neutron absorber 10 is shown and described herein as it may comprise a part or portion of a test cell 12 that is designed to expose sample material 14 to a fast-flux neutron field. More specifically, and in the embodiments shown and described herein, the test cell 12 is configured to be used in conjunction with the Advanced Test Reactor located near Idaho Falls, Id., and operated by the U.S. Department of Energy's Idaho National Laboratory. Basically, the Advanced Test Reactor comprises a “four leaf clover” core design that provides a plurality of test spaces or regions suitable for receiving materials to be tested. The Advanced Test Reactor is capable of producing neutron fields having extreme neutron fluxes, thereby allowing materials to be exposed in a few weeks or months to neutron radiation that would take years in a typical commercial reactor. Because the particular embodiments shown and described herein are intended to be used in the Advanced Test Reactor, certain aspects (e.g., the overall size and shape) of the test cell 12, as well as the conduction cooled neutron absorber 10 contained therein, are based on requirements imposed by the particular test reactor, as opposed to any special requirements of using the invention. However, the particular requirements imposed by the desired use with the Advanced Test Reactor will become readily apparent to persons having ordinary skill in the art after having become familiar with the teachings provided herein. Consequently, the present invention should not be regarded as limited to conduction cooled neutron absorbers having the particular sizes, configurations, or other design aspects specified herein that are required for the particular application. In accordance with the desired use of the conduction cooled neutron absorber 10 in the Advanced Test Reactor, the conduction cooled neutron absorber 10 comprises a generally cylindrically shaped bulk material structure or heat sink portion 16 of the test cell 12. The conduction cooled neutron absorber 10 may be provided with one or more bores 18 therein for receiving the sample or test material 14, as well as one or more coolant channels 20 for receiving coolant 22. In one embodiment, the conduction cooled neutron absorber 10 is surrounded or encapsulated by a pressure tube 30. Pressure tube 30 allows the pressure exerted by the pressurized coolant 22 flowing in coolant channels 20 to be borne by the pressure tube 30 rather than by the conduction cooled neutron absorber 10. In addition, encapsulation of the conduction cooled neutron absorber 10 by the pressure tube 30 allows the test cell 12 to be more easily handled as a self-contained component that can be readily placed within the test reactor. Pressure tube 30 may comprise any of a wide range of materials and have any of a wide range of thicknesses suitable for the intended application, as would become apparent to persons having ordinary skill in the art after having become familiar with the teachings provided herein. Consequently, the present invention should not be regarded as limited to a pressure tube 30 comprising any particular material or thickness. However, by way of example, in one embodiment, the pressure tube 30 is fabricated from INCONEL® 600 and has a thickness of about 3.175 mm (about 0.125 inch). in the particular embodiments shown and described herein, the cylindrically shaped bulk material structure 16 of the conduction cooled neutron absorber 10 may have an overall diameter 32 of about 7.3 cm (about 3.386 inches) and an overall length 34 of about 1.22 m (about 48 inches). Each of the bores 18 may have a diameter of about 3.17 cm (about 1.25 inches) and may be arranged generally evenly (e.g., at intervals of 120°) around the geometric center of the cylindrically shaped structure 16. Alternatively, other configurations are possible, some of which are shown and described herein and others of which will become apparent to persons having ordinary skill in the art after having become familiar with the teachings provided herein. Consequently, the present invention should not be regarded as limited to conduction cooled neutron absorbers having any particular configuration. As already mentioned, the conduction cooled neutron absorber 10 may also be provided with one or more coolant channels 20 therein through which are caused to flow a coolant 22. Coolant 22 removes heat resulting from the neutron radiation (e.g., heat resulting from the neutron irradiation of the sample material 14 as well as from the absorption of neutrons by the conduction cooled neutron absorber 10). In the embodiments shown and described herein, the coolant 22 may comprise water, although other coolants, such as molten salts or liquid metals, may also be used. In the first embodiment, four coolant channels 20 of conduction cooled neutron absorber 10 are provided at the locations best seen in FIGS. 1 and 2. By way of example, each of the coolant channels 20 may have a diameter of about 6.35 mm (about 0.25 inch). Alternatively, other arrangements and configurations are possible, as will be described in greater detail below. The sample material 14 may be provided within the various bores 18 provided in the conduction cooled neutron absorber 10. Any of a wide variety of configurations and arrangements are possible for providing the sample material 14 in the bores 18, some of which are described herein and others of which will become apparent to persons having ordinary skill in the art after having become familiar with the teachings provided herein. Consequently, the present invention should not be regarded as limited to any particular configuration and arrangement for providing the sample material 14 within the bores 18. However, by way of example, in one embodiment, the sample material 14 may be surrounded by a jacket 36. Depending on the particular application and the heat loads expected, a thermal bonding agent 38 may be provided between the material 14 and the jacket 36 in order to improve the heat transfer between the material 14 and jacket 36. Jacket 36 may comprise any of a wide range of materials suitable for the intended application, as would become apparent to persons having ordinary skill in the art after having become familiar with the teachings provided herein. Consequently, the present invention should not be regarded as limited to any particular jacket 36 (or even to the presence of such a jacket 36). However, by way of example, in one embodiment, the jacket 36 may comprise INCONEL® 600. Examples of suitable thermal bonding agents 38 include liquid metals or salts, although other materials may be used. It may also be desirable to size the overall sample assembly 40 (i.e., the sample material 14, jacket 36, and thermal bonding agent 38) so that an annulus 42 is defined between the jacket 36 of sample assembly 40 and bore 18. Annulus 42 may receive a temperature control gas (not shown), such as a mixture of helium and neon, to act as a thermal interface between the sample material 14 and the conduction cooled neutron absorber 10. Alternatively, such an annulus 42 need not be provided. Instead, the jacket 36 could be sized to contact some or all of the interior surface of bore 18 to provide direct contact for heat transfer, as will be described in further detail below. Before proceeding with the description, it should be noted that the particular configuration of the sample assembly 40, i.e., whether the sample material 14 will be provided with a jacket 36, a thermal bonding agent 38, and/or be sized so as to define an annulus 42 between the jacket 36 and bore 18 will depend on a wide range of factors, such as, for example, the type of sample material 14 that is to be used, the particular heat loads expected, the coefficient of thermal expansion of the sample material 14, the particular coolant 22 to be used, and other factors, as would become apparent to persons having ordinary skill in the art after having become familiar with the teachings provided herein. Moreover, because the particular configuration and structural arrangement of the sample assembly 40 is not particularly critical in achieving the various objects and advantages of the present invention, the structure of the particular sample assembly 40 that may be utilized in conjunction with the present invention will not be described in further detail herein. The conduction cooled neutron absorber 10 is formed from a composite material that comprises various constituents that provide the conduction cooled neutron absorber 10 with various advantageous properties and characteristics. More specifically, the composite material may comprise a first material or constituent component having a large thermal neutron cross-section, such as for example, a thermal neutron cross-section of at least about 50 barns. This first material or constituent component is dispersed within a second material or constituent component having a high thermal conductivity (e.g., at least about 1 W/cm·K) so that a substantially homogenous mixture of the two materials is formed. Ideally, the particular constituents of the composite material should be stable (e.g., not subject to transmutation) in the expected nuclear environment and should be capable of being formed into the desired shape. Of course, the two materials should be generally non-reactive with one another, so that the resulting composite material is stable. A suitable material for the first constituent component is hafnium. Hafnium is a good thermal neutron absorber, having a thermal neutron cross-section of about 104 barns. In addition, hafnium is known to be generally stable in the reactor environment. A suitable material for the second constituent component is aluminum. Aluminum has a thermal conductivity of about 2.37 W/cm·K and is also known to be generally stable in the nuclear environment. Generally speaking, hafnium is not soluble in aluminum. However, various hafnium/aluminum intermetallic phases are soluble in aluminum. Of the various hafnium/aluminum intermetallic phases that may be formed, it is generally preferred to utilize an intermetallic phase having the least amount of hafnium, specifically, hafnium aluminide (Al3Hf). Compositions comprising hafnium aluminide (Al3Hf) in aluminum are stable up to about 25 atomic percent hafnium and up to the melting temperature of aluminum (i.e., about 660° C.). The amount of hafnium that should be provided in the aluminum may be varied depending on the particular requirements of the system (e.g., on the desired fast-to-thermal ratio of neutrons to be provided), as well as on the particular configuration of the conduction cooled neutron absorber 10. The amount of hafnium that should be provided in a particular design may be determined with the aid of any of a wide range of computer models or simulations. By way of example, in one embodiment, the Monte-Carlo Neutronics-Particle Transport Code known as MCNP (version 1.40) was used to model proposed designs and to provide estimates of the fast-to-thermal neutron ratios that may be achieved by incorporating various amounts of hafnium into the composite neutron absorber material. Briefly, the MCNP computer code is a general purpose Monte-Carlo N-Particle Transport Code that can be used for neutron, photon, electron, or coupled neutron/photon/electron transport modeling. The MCNP computer code is well known in the art and is available from Los Alamos National Laboratory. Referring now primarily to FIG. 3, with continued reference to FIGS. 1 and 2, the MCNP computer code was used to model a conduction cooled neutron absorber 10 having three bores 18 therein for receiving the sample material 14. The MCNP computer code was then used to generate a plot of the fast-to-thermal neutron ratio and fast flux intensity that would exist in the bores 18 for hafnium/aluminum composite materials comprising various amounts of hafnium. For example, and as illustrated in FIG. 3, the conduction cooled neutron absorber 10 will perform well and achieve fast-to-thermal neutron ratios of at least about 50 when hafnium is present in aluminum up to about 10 atomic percent. Fast-to-thermal neutron ratios of about 40 can be achieved with hafnium levels of about 7 atomic percent. These amounts of hafnium are well within the stability range of hafnium aluminide in aluminum, as described above. Heating rates of the hafnium/aluminum composite material were also modeled and are depicted in FIG. 4. The heating rate, which is a measure of thermal neutron absorption, appears to saturate at about 6-7 atomic percent hafnium, which suggests that hafnium levels of about 6-7 atomic percent are optimum. It should be noted that the computer simulations depicted in FIGS. 3 and 4 were for a “dry” neutron absorber, i.e., with no coolant channels 20 and no water present as coolant 22. However, and as illustrated in FIG. 5, additional computer analysis with models involving water in various configurations of coolant channels 20 indicates that the system is primarily sensitive to the total amount of water coolant in the system, rather than on the particular arrangement of the coolant channels 20. As already mentioned, hafnium is not significantly soluble in aluminum. Consequently, a conduction cooled neutron absorber 10 comprising hafnium and aluminum is not amenable to formation by conventional alloying techniques. Consequently, such hafnium/aluminum composite materials must be formed by other techniques. In one embodiment, the hafnium/aluminum composite material is formed by mixing together powders containing hafnium and aluminum powders. The resulting powder mixture may then be consolidated to form the conduction cooled neutron absorber 10. More specifically, and with reference to FIG. 6, a conduction cooled neutron absorber 10 comprising the aluminum/hafnium composite material described herein may be produced by mixing together a hafnium aluminide intermetallic powder 24 (e.g., hafnium aluminide powder, (Al3Hf)) with aluminum powder 26 form a powder mixture 28. The amount of hafnium to be provided in the final conduction cooled neutron absorber 10 product may be controlled by varying the amount of hafnium aluminide intermetallic powder 24 (e.g., Al3Hf) that is mixed with the aluminum powder 26. For example, in order to achieve the atomic percentages of hafnium described herein, the powder mixture 28 may comprise from about 10 volume percent to about 50 volume percent of the hafnium aluminide intermetallic powder 24 with the balance (i.e., from about 90 volume percent to about 50 volume percent) comprising the aluminum powder 26. A hafnium concentration of about 7 atomic percent, wherein the hafnium is provided as Al3Hf, may be achieved with a powder mixture 28 containing about 23 volume percent hafnium aluminide intermetallic powder 24 (e.g., Al3Hf), with the balance (i.e., about 77 volume percent) comprising aluminum powder 26. The hafnium aluminide intermetallic powder 24 (e.g., Al3Hf) and aluminum powder 26 may be provided in any of a wide range of powder sizes suitable for the particular powder compaction processes that may be utilized to form the conduction cooled neutron absorber 10. Consequently, the present invention should not be regarded as limited to powders 24, 26 having any particular sizes or within any particular range of sizes. However, by way of example, the hafnium aluminide intermetallic powder 24 may be provided having particle sizes in the range of about 45 μm to about 200 μm (i.e., −70 +325 U.S. Tyler mesh). The aluminum powder 26 may be provided having a particle of about 45 nm or less (i.e., −325 U.S. Tyler mesh). However, it should be noted that density differences between the powders 24 and 26 may result in the separation (i.e., unmixing) of the resulting powder mixture 28 if suitable precautions are not taken. For example, hafnium aluminide (Al3Hf) is more than twice as dense as aluminum, so it may be desirable to provide the Al3Hf in a larger particle size compared to the particle size of the aluminum. The size differential of the particles may reduce the tendency of the particles to separate due to differences in density. Alternatively, other methods may be used to ensure that the hafnium is generally homogenously dispersed within the aluminum in the final conduction cooled neutron absorber 10 product. Once a suitable powder mixture 28 has been produced, the powder mixture 28 may then be consolidated to form the neutron absorber 10 having the desired configuration. Any of a wide range of consolidation processes or techniques that are now known in the art or that may be developed in the future may be used to consolidate the powder mixture 28 into the desired configuration. For example, the powder mixture 28 may be consolidated by any of a wide range of well-known compaction or hot isostatic pressing processes. Alternatively, the powder mixture 28 may be consolidated by extruding. If desired, the resulting compact may be further consolidated and/or densified by subjecting it to a subsequent sintering step. Depending on the desired configuration, the conduction cooled neutron absorber 10 may be formed in a single consolidation step. Alternatively, the conduction cooled neutron absorber 10 may be formed by joining together (e.g., by pressure welding) separately formed pieces or components. Various configurations for the conduction cooled neutron absorber 10 may be modeled in advance of fabrication by utilizing any of a wide range of computer modeling and simulation techniques to test proposed design configurations. For example, a thermal/hydraulic analysis of a proposed design may be conducted by using a finite element analysis computer code, such as ABAQUS® (e.g., version 6.6-3). The ABAQUS® computer code is well known in the art and is available from Dassault Systems, Inc., of Providence, R.I. (US). Referring now primarily to FIG. 7, an example configuration of a conduction cooled neutron absorber was modeled to investigate the heat loads and temperature distributions that may be expected in use. More specifically, the computer model was based on the configuration illustrated in FIG. 7, wherein the conduction cooled neutron absorber 10 defines three specimen chambers or bores 18 with four coolant channels 20 arranged in the manner illustrated in FIG. 7. The conduction cooled neutron absorber 10 is surrounded by a pressure tube 30 and an envelope tube 44. A cooling annulus 46 is defined between the pressure tube 30 and the neutron absorber 10, whereas a second annulus 48 is defined between the envelope tube 44 and the pressure tube 30. Second annulus 48 may be filled with helium and is used for the purposes of leak detection. More specifically, in the configuration modeled, an outside diameter of the conduction cooled neutron absorber 10 is 7.229 cm (3.386 inches). The pressure tube 30 has an inside diameter of 8.753 cm (3.446 inches) and an outside diameter of 9.388 cm (3.696 inches). The envelope tube 44 has an inside diameter of 9.591 cm (3.776 inches) and an outside diameter of 10.226 cm (4.026 inches). The experiment tubes or bores 18 each have a diameter of 3.17 cm (1.25 inches) and are arranged on circle 50 having a diameter of 4.56 cm (1.795 inches). Each of the coolant channels 20 has a diameter of 6.35 mm (0.25 inch). Three of the coolant channels 20 are located on a 3.09 cm (1.22 inches) circle 52, with the fourth coolant channel 20 being located at the geometric center of the conduction cooled neutron absorber 10. The coolant annulus 46 has a thickness of 0.762 mm (0.030 inch), whereas the second or leak detection annulus 48 has a thickness of 1.016 mm (0.040 inch). For modeling purposes, all components (i.e., the conduction cooled neutron absorber 10, pressure tube 30 and envelope tube 44) have overall lengths of 1.219 m (48 inches). The conduction cooled neutron absorber 10 comprises hafnium/aluminum composite material comprising about 7 atomic percent hafnium. Both the pressure tube 30 and envelope tube 44 comprise INCONEL® 600. In the model, coolant water to the cooling annulus 46 and the coolant channels 20 is provided at a pressure of about 12.4 MPa (1800 psia) and has an inlet temperature of 110° C. (230° F.). The flow rate of the water coolant through the coolant annulus 46 was selected to be 90 L/m (23.8 gpm), whereas the rate through the coolant channels 20 was set at 116.9 L/m (30.9) gpm. The cooling annulus 46 and water channel 20 mass flow rates were specified as 0.687 kg/s·cm2 (9.76 lbm/s·in2) and 1.464 kg/s·cm2 (20.81 lbm/s·in2), respectively. The total volume of water is 4.076×10−4 m3 (24.87 in3), with 0.2532 L (15.45 in3) in the cooling annulus 46 and 0.1544 L (9.422 in3) in the four water channels 20. Oxide layers having a thickness of 12.7 μm (0.0005 inch) are assumed to be present on the surfaces of the conduction cooled neutron absorber 10 that are exposed to the coolant 22 which, in this example, comprised water. Another computer modeling program, known as “RELAP5-3d,” was used to specify the convection boundary conditions at the outside of the envelope tube 44 in order to simulate conditions in the Advanced Test Reactor. Briefly, RELAP5-3d is a computer simulation code that may be used to perform thermal/hydraulic analyses and kinetic modeling of nuclear reactor systems. RELAP5-3d is well known in the art and is available from the Idaho National Laboratory of Idaho Falls, Id. (US). The external surface of envelope tube 44 is cooled by the primary coolant (water) of the Advanced Test Reactor flowing at a velocity of 13.59 m/s (44.6 ft/s) with an average coolant temperature of 82.2° C. (180° F.). To facilitate specification of heat loads, the geometry of the conduction cooled neutron absorber 10 was partitioned into twelve axial sections with six 10.16 cm (4 inch) sections above and below the core mid-plane. FIG. 8 is a perspective view of a computer-generated isothermal plot showing the internal temperature distribution of the conduction cooled neutron absorber 10 contained in the test cell configuration of FIG. 7. The temperature in the conduction cooled neutron absorber 10 reaches a maximum of about 224.2° C. (about 435.6° F.) in the regions between the bores 18 at about the center of the envelope tube 44, i.e., about 61 cm (about 24 inches) from either end of the conduction cooled neutron absorber 10. The calculated average heat loads for each component modeled are presented in Table 1: TABLE 1AverageHeat FluxComponentHeat Load (kW)ConditionExperiment Tubes70 each/210 totalSurfaceNeutron Absorber40.3BodyPressure Tube58.32BodyEnvelope Tube70.5BodyWater Channels3.741BodyWater Annulus6.135Body The computer water channel outlet temperatures are presented in Table 2: TABLE 2LocationTmax (° C./° F.)Water Annulus142.7/288.8Center Water Channel135.6/276.1Off-Center Water Channels (3)124.9/256.8 In summation, the proposed design of the conduction cooled neutron absorber 10 is capable of maintaining all system components below their maximum temperature limits. The maximum temperature of the conduction cooled neutron absorber 10 is about 224.2° C. (435.6° F.) and occurs in a small, localized region of the conduction cooled neutron absorber 10 near the core mid-plane (e.g., about 61 cm (about 24 inches)) from either end of the neutron absorber 10, which is well below the melting temperature of aluminum (about 660° C.). The total coolant flow rate requirement for the modeled configuration is about 207 L/m (about 54.7 gpm). If desired, the design could be further refined to reduce the amount of water coolant required, which should serve to increase the fast-to-thermal neutron ratio in the experiment locations (e.g., within bores 18). As mentioned above, other variations are possible and may also be used. For example, a second embodiment of a conduction cooled neutron absorber 110 is illustrated in FIG. 9 and differs from the conduction cooled neutron absorber 10 in that the conduction cooled neutron absorber 110 is not provided with any coolant channels therein. Instead, the conduction cooled neutron absorber 110 is provided with a cooling annulus 146 adjacent a pressure tube 130. Sample material assemblies 140 containing sample material 114 may be provided in various bores 118 provided in the conduction cooled neutron absorber 110. In the embodiment illustrated in FIG. 9, the sample material 114 may be surrounded by a jacket 136 and a thermal bonding agent 138 in the manner already described for the first embodiment. The jacket 136 may be sized so that an annulus 142 is created between the sample material assemblies 140 and the bores 118. Annulus 142 may receive a temperature control gas (not shown), such as a mixture of helium and neon, to act as a thermal interface between the sample material 114 and the conduction cooled neutron absorber 110. Alternatively, such an annulus 142 need not be provided. A third embodiment of a conduction cooled neutron absorber 210 is illustrated in FIG. 10 and may instead utilize three high aspect ratio cooling channels 220 arranged in a “Y” configuration in the manner illustrated in FIG. 10. Note that a separate annular cooling channel is not provided between the conduction cooled neutron absorber 210 and pressure tube 230, although such an arrangement is possible. Sample material assemblies 240 may comprise a structural arrangement basically identical to those already described. That is, each sample material assembly 240 may comprise a quantity of sample material 214 surrounded by a jacket 236 with a thermal bonding agent 238 therebetween. The sample material assemblies 240 may be sized so that an annulus 242 is defined between the sample material assemblies 240 and bores 218 provided in the conduction cooled neutron absorber 210. A temperature control gas (not shown), such as a mixture of helium and neon, may be caused to flow through annulus 242 during operation. Still yet other variations and configurations of the conduction cooled neutron absorber are possible. For example, a fourth embodiment of a conduction cooled neutron absorber 310 illustrated in FIG. 11 may comprise a plurality of smaller coolant channels 320 provided in the arrangement shown in FIG. 11. Like the embodiment illustrated in FIG. 10, an annular coolant channel is not provided adjacent a pressure tube 330, although such a coolant annulus could be provided. In a departure from the other embodiments already described, a sample material 314 of the fourth embodiment of the conduction cooled neutron absorber 310 may be provided within a star-shaped jacket 336 that is sized to contact the interior surface of bore 318. The star-shaped jacket 336 thereby provides a direct heat path (e.g., by contact) between the material 314 and the conduction cooled neutron absorber 310. A thermal bonding agent 338 may be provided between sample material 314 and jacket 336. A temperature control gas (e.g., a helium/neon mixture) may be caused to flow in spaces 342 between the star-shaped jacket 336 and bore 318, if desired. Having herein set forth preferred embodiments of the present invention, it is anticipated that suitable modifications can be made thereto which will nonetheless remain within the scope of the invention. The invention shall therefore only be construed in accordance with the following claims. |
|
summary | ||
abstract | A method for evaluating the clogging of the passages of a tube support plate of a tube heat exchanger, in which an eddy current probe is passed through a tube of the exchanger and a measurement signal is measured with the probe to evaluate the clogging at the downstream edge of a tube support plate. A lower (upper, respectively) edge signal corresponding to the probe passing the downstream (upstream, respectively) edge of the tube support plate is determined from the measurement signal. The impulse response of the probe is estimated. The lower edge signal is deconvolved (530) by the impulse response estimation. The clogging is evaluated (533) by analyzing (532) the lower edge signal thus deconvolved. |
|
050858254 | description | DETAILED DESCRIPTION OF THE INVENTION Referring to FIG. 1 of the drawing, a water cooled nuclear fission reactor plant 10 includes a reactor pressure vessel 12, containing a core of fissionable fuel material 14, such as enriched uranium oxide pellets contained within sealed metal tubes grouped into conveniently sized bundles. The bundles are arranged in a same-orientation pattern generally comprising collectively as a right circular cylinder of nominally comparable cylindrical length and diameter dimensions. During reactor power generation operations, the neutron incited fission reaction of the fuel material is controlled by neutron-absorbent control rods or blades positioned into appropriate amounts and patterns of control rod withdrawal to produce prodigious amounts of thermal energy. The core of fuel assemblies is substantially submerged in water coolant 16 which circulates through the fuel core 14 to carry away heat and form steam for work, such as driving a turbine for generating electrical power. Control rods 18 containing a neutron absorbing material such boron, are reciprocally moveable into and out from the fuel core 14 to control or govern the rate of the neutron incited fission reaction of the fuel, or to terminate the reaction. This in turn safely regulates the quantities of heat produced for work. Typically such nuclear reactor plants 10 are provided with an auxiliary cooling water system 20, including a supplementary cooling water reservoir tank 22 or vessel containing a standby supply of supplementary water coolant 24. Tank 22 can be resupplied with water 24 through fluid conduit 26 from a suitable source (not shown). Cooling water auxiliary systems 20 are commonly charged with a propellant gas under pressure, such as nitrogen, as a means for driving supplementary water coolant 24 from a reservoir tank 22 to the reactor vessel 12. A propellant gas is provided for supplementary coolant reservoir tank 22 from high pressure gas supply tank 28 through fluid conduit 30. Thus supplementary coolant water 24 can be forced by the pressurized gas propellant from reservoir tank 22 through fluid conduit 32 and injected into the reactor vessel 12 to replace the operating reactor coolant water 16 covering the fuel core 14 when that coolant is abruptly lost due to a major breach in the coolant circulating system. The transfer of supplementary coolant water 24 from reservoir tank 22 through conduit 32 to the reactor vessel 12 is controlled by a flow injection valve 34 provided in conduit 32 which opens in response to a signal from a sensing device 36 located within the reactor circulating system which is passed to a valve actuating means 38 operating flow injection valve 34. Commonly these nuclear reactor plants 10 are also provided with an emergency neutron absorption system 40, comprising a supply of a neutron absorbing water solution 42, such as a water soluble boron compound, held ready in a supply tank 44 or vessel. Tank 44 can be resupplied with the neutron absorbing aqueous solution through fluid conduit 48 from a suitable source (not shown). Emergency neutron absorbing systems 40 are also commonly charged with a propellant gas under pressure, such as nitrogen, as a means for driving the emergency neutron absorbing solution 42 from the supply tank 44 to the reactor vessel 12. The propellant gas can be provided to the neutron absorption solution supply tank 44 from high pressure gas supply tank 28 through gas conduit 46. Thus emergency neutron absorbing solution 42 can be forced by pressurized gas propellant from supply tank 44 though fluid conduit 50 and injected into the reactor vessel 12 to envelop the fuel core 14 with neutron absorbing liquid in the unlikely event the normal operational injection control rods malfunction. The transfer of emergency neutron absorbing solution 42 from supply tank 44 through conduit 50 to the reactor vessel 12 is controlled by a flow injection valve 52 provided in conduit 50. Valve 52 opens in response to a signal from a sensing device 54 monitoring the operation of the control rods 18 which is passed to a valve actuating mechanism 56 operating flow injection valve 52. A pressure sensing device 58 monitors the propelling gas pressure within the supplementary cooling water reserve tank 22 and controls a valve actuating mechanism 60 which operates valve 62 in gas supply pipe 30 communicating with high pressure gas supply tank 28. Thus, the pressure of the propelling gas in tank 22 for forcing cooling water 24 when needed from reservoir tank 22 through fluid conduit 32 into the reactor vessel 12 can automatically be maintained within an appropriate range by suitable additions of pressurized gas from the high pressure gas supply tank 28. Similarly a pressure sensing device 64 monitors the propelling gas pressure within the emergency neutron absorbing solution supply tank 44 and controls a valve actuating mechanism 66 which operates valve 68 in pipe 46 communicating with high pressure gas supply tank 28. The pressure of the propelling gas in tank 44 for forcing neutron absorbing solution 42 when needed from reservoir tank 44 through fluid conduit 50 into the reactor vessel 12 accordingly can automatically be maintained within an appropriate range by suitable additions of pressurized gas from the high pressure gas supply tank 28. In the event the gas pressure within the high pressure gas supply tank 28 is insufficient to pressurize by passive flow action either tank 44 or tank 22 up to its required initial operating pressure, it is routine to provide a gas compressor of suitable design and driving means to accomplish the charging task. Typically the propellant gas pressure is maintained at different levels, or ranges, in the supplementary water coolant tank 22 and neutron absorption solution supply tank 44, and these tanks are of substantially different size or volume to correspond with their respective functions. The supplementary water coolant tank 22 normally would be of much greater size for the purpose of containing a large reserve volume of water coolant 24 to supplant significant losses from the reactor of reactor coolant 16 over a period of time, whereas the neutron absorbent solution 42 is generally needed only in an amount sufficient to produce a dilute but effective concentration within coolant 16 of neutron absorbing solution and thus to bring fuel core 14 subcritical in the absence of control rod action. The propellant gas pressure within the supplementary water coolant tank 22 is relatively low since it is generally required to drive the supplemental water coolant into a reactor system which is losing pressurization and coolant due to a breach, for example about 400 pounds per square inch gauge. However, to maintain needed effectivity of the initial propellant gas within tank 22 while reserve coolant 24 is being discharged through conduit 32 and to the point of depletion within tank 22, the physical size of the initial gas volume with tank 22 is necessarily relatively large, such as being one or more multiples (plus/minus) of the coolant 24 initial volume. The consequence of this is that tank 22 can in the practice become a bulky heavy tank presenting onerous problems concerning its manufacturing, shipping, installation, and space accommodation within costly plant building structure among other problems which attend such bulky heavy equipment. Propellant gas pressure within neutron absorption solution supply tank 44 is relatively high so that it can drive the solution into the reactor vessel 12 even under scenario in which reactor pressure develops to become untypically high, for example about 1500 pounds per square inch gauge. According to this invention the propellant pressurized gas supply and gas distribution means associated with tanks 22 and 44 are integrated by interconnecting the pressurized supplementary water coolant tank 22 with the pressurized emergency neutron absorption solution supply tank 44 for the purpose of transferring pressurized gas, suitably conditioned in pressure directly between tanks 44 to 22. Referring to FIG. 1, a conduit 70 makes fluid communication directly between the upper gas containing area of emergency neutron absorption solution tank 44 and supplementary water coolant tank 22. Fluid flow through conduit 70 is regulated by flow control valve 72. Flow control valve 72 is operated by valve actuating mechanism and pressure monitoring and sensing device 74 which measures the pressure in both tanks 22 and 44. Pressure monitoring devices 74, for example, can be programmed to actuate flow control valve 72 and permit the flow of pressurized gas from one of the tanks 44 or 22 to the other tank when the pressurization of one tank is reduced to a predetermined low level, due either to its dissipation from driving liquid to the reactor vessel 12, or simply leakage. This arrangement of the invention provides a significant safety factor by providing an added source or reservoir of standby pressurized gas propellant for effectively driving either the supplementary coolant water to the reactor vessel for maintaining moderate temperature, or the emergency neutron absorbing solution to the reactor fuel core for terminating the fission reaction. In addition to enhancing safety, this arrangement affords economics in the plant facilities such as the size of reservoir tanks for containing the supplementary coolant water and pressurized gas propellant, or the emergency neutron absorbent solution and pressurized gas propellant. This can be attributable to reducing the tank volume or size utilized to contain the pressurized gas propellant in the area above the liquid since an additional standby source pressurized gas propellant is available to augment the gas content of either tank. Another advantageous aspect of this invention is an automatic cut-off arrangement which precludes the introduction of gas propellant into the reactor vessel 12 though either the auxiliary water coolant system 20 or the emergency neutron absorption solution system 40. A liquid level measuring system is applied to each the supplementary water coolant tank 22 and the neutron absorption solution supply tank 44 and provided with control means to terminate all fluid flow from these tanks to the reactor vessel 12 upon a reduction of their liquid content to a predetermined minimum level therein. Supplementary water coolant tank 22 is provided with a liquid level sensor 76 which registers a minimum liquid level within tank 44. Sensor 76 regulates valve actuating mechanism(s) 78 (78') which operates flow control valve(s) 80 (80'). Thus, when the water coolant in tank 22 is reduced by discharge to a predetermined level, flow control valve(s) 80 (80') closes to prevent flow though coolant pipe 32 from tank 22 to the reactor vessel 12. Multiple flow control valves 80 can be used to insure positive performance through duplication of the units. Similarly neutron absorption solution supply tank 44 is provided with a liquid level sensor 82 which registers a minimum liquid level within tank 44. Sensor 82 regulates valve actuating mechanism(s) 84 which operate flow control valve(s) 86. Accordingly when the neutron absorbent solution tank 44 is reduced to a predetermined level, flow control valve(s) 86 closes to prevent flow through solution pipe 50 from tank 44 to the reactor vessel 12. Also multiple flow control valves 86 can be used to insure positive performance, such as the 10 valves 80 and 80' in supplementary coolant pipe 32 and valve control mechanisms 78 and 78'. To further insure that any gas propellant does not escape from either supplementary water coolant tank 22 or neutron absorption solution supply tank 44 and enter the reactor, optionally each tank is provided with a float operated valve 88 and 88', housed within a cage, having a valve means for descending down to and closing off the bottom tank outlets when their liquid contents recede to given low point. To preclude any possibility of radioactive contaminated coolant water circulating about through the nuclear reactor vessel 12 and coolant/steam circuit from flowing back through supplementary water coolant injection pipe 32, a check valve 90 is preferably provided at the discharge end of pipe 32 within reactor vessel 12. The portion of conduit 32 downstream (i.e., on the reactor-side) of squib valves 34 is typically pressurized internally to reactor operating pressures. Accordingly, this pipe segment is included among that group of pipe segments and appurtenances to reactor 12 the rupture of which (segments/appurtenances) are one-at-a-time taken as hypothetical initiating events which, together with other presumed happenstances, comprise the spectrum of accident scenarios for which coolant injection system 20 is intended to provide mitigating relief. Rupture of the subject pipe segment along conduit 32 could thus represent the initiating event in the hypothetical accident scenario, which in turn would foreclose the possibility for coolant 24 in tank 22 from reaching reactor 12. To address the need by the reactor for coolant injection irrespective of the initiating event, heretofore it has been commonplace in the art for the reactor designer to provide redundant coolant injection systems 20 each with dedicated injection lines 32, 32'. In many cases, this has led to large extra costs for these backup systems, and to extra and costly demands within the reactor auxiliary building for space to house the backup components. An additional novel feature of this invention is now herewith disclosed, depicted in FIG. 2, being the incorporation of a removable check valve 90, normally-closed, positioned at the in-reactor terminus of a "safe-end" 92, which latter object is in the art understood to be an inward-to-the-reactor projecting pipelike segment communicating with an external line such as line 32, and suitably terminated in the reactor vessel with a removable check valve 90 affixed to safe-end 92 via clamp 94. Check valve 90 affords near-zero flow resistance to inwardly directed flow coming through safe-end 92, but presents virtually infinite flow resistance to any outwardly directed flow moving through safe-end 92. Rupture of safe-end 92 is ordinarily not held to be a credible initiating event because of the fact that pressure within and without safe-end 92 is normally identical or nearly so; accordingly, mechanical stresses are ordinarily near-zero, and thus catastrophic rupture is not credible. Rupture of line 32 is, however, a credible initiating event; but with line 32 communicated with inside-the-reactor safe-end 32 and check valve 90, it is at once evident that catastrophic rupture of line 32 is inherently promptly mitigated by the action of check valve 90 in its normally-closed state, limiting leakage of reactor coolant outwardly through line 32 to zero or to nearly-zero values and thus generally obviating the need for reactor safety coolant injection systems--such as coolant injection system 20--to have to function. Instead, ordinary reactor shutdown and depressurization actions are taken by the plant operator or the nuclear reactor's automatic systems upon the detection of the hypothesized catastrophic rupture (=initiating event). Check valve 90 is not required to be leak-tight, but only to be capable of limiting outleakage to a low flow rate the specific amount of which being nominally that of the reactor's ordinary high-pressure coolant makeup system assuming unavailability of feedwater (this assumption typically being part of the hypothetical accident scenario of events). By utilizing the novel combination of safe-end 92 and normally-closed check valve 90 affixed thereto with clamp 94, it becomes possible to derive the otherwise unanticipated benefit of eliminating the backup safety coolant injection system for nuclear reactor plants the design basis accidents for which involve no greater than one catastrophic rupture of a pressurized pipeline segment together with any other single active failure. It is also preferred in this invention that flow control valves 34 and 52, respectively in supplementary coolant pipe 32 leading from tank 22 to vessel 12 and in emergency solution pipe 50 leading from tank 44 to vessel, consist of multiple parallel valves such as the pairs of valves, 34 and 52, arranged in parallel as shown in the drawing. Additionally it is preferred that the valves 34 and 52, either single or in parallel pairs, be of the explosive charge type whereby an explosive charge functions to remove the valve closure blocking flow through the coolant injection pipe 26 and the solution injection pipe 50. Additionally a squib-type, or explosive activated valve 96 is preferably located in conduit 70 which opens in response to a logic system signal as a means of providing a highly reliable non-leak isolating action of the high pressure gas in tank 44 from bleeding down into low pressure tank 22. In this embodiment, sensor means 36 and 54 can be used together with appropriate logic systems to initiate a valve-operating means 98 to open valve 96. Also valve 96 can be provided as dual valves in parallel such as shown for valve 34 in the drawing. |
053176154 | abstract | An exposure apparatus and method for exposing a workpiece to a pattern of an original with radiation includes a masking device having movable blades for variably defining an aperture to selectively block and transmit the radiation to define on the workpiece a desired exposure zone corresponding to the aperture, the masking device having a window, and a detector for detecting the positional deviation between the original and the workpiece by using light passing through the window of the masking device. |
description | Pursuant to 35 U.S.C. § 119(a), this application claims the benefit of the earlier filing date and the right of priority to Korean Patent Applications No. 10-2017-0150719, filed on Nov. 13, 2017, the contents of which are incorporated by reference herein in their entirety. The present disclosure relates to a safety system for securing the safety of a nuclear power plant, and more particularly, to a facility capable of reducing a pressure inside a containment and a concentration of radioactive material by a passive principle when a nuclear accident occurs. Nuclear reactors are divided into loop type reactors (e.g., commercial reactors: Korea) in which major components (steam generator, pressurizer, pump, etc.) are installed outside a reactor vessel and integral reactors (e.g., SMART reactors: Korea) in which the major components are installed inside a reactor vessel. In addition, reactors are divided into active reactors and passive reactors depending on the implementation method of a safety system. An active reactor is a reactor using an active component such as a pump operated by electric power of an emergency generator or the like to drive a safety system, and a passive reactor is a reactor using a passive component operated by gravity, gas pressure or the like to drive a safety system. A passive safety system in a passive reactor may maintain the reactor in a safe manner only with a natural force built in the system without an operator action or an AC power source of safety grade such as an emergency diesel generator for more than a period of time (72 hours) required by regulatory requirements in the event of an accident, and the safety system may receive assistance from an operator action or a non-safety system after 72 hours. A containment (containment building, reactor building, containment vessel, safeguard vessel, or the like) serving as a final barrier to prevent leakage of radioactive material from a reactor to an external environment may be divided into a containment building (or referred to as a reactor building) formed of reinforced concrete, a containment vessel and a safeguard vessel formed with a steel vessel depending on a material constituting a pressure boundary thereof. The containment vessel is a large vessel designed at low pressure such as the containment building, and the safeguard vessel is a small vessel designed to be small by increasing the design pressure. Unless otherwise specified, the containment building, the reactor building, the containment vessel or the safeguard vessel are collectively referred to as a containment. For a method of reducing the pressure and temperature inside the containment and reducing the concentration of radioactive material, active and passive systems are used in various forms such as a containment spray system, a containment cooling system, a suppression tank or a suppression sump. Hereinafter, these facilities will be sequentially described below. The active containment spray system (Korea commercial reactor, SMART reactor, etc.) method performs a function of spraying a large amount of cooling water using a pump during an accident, collecting cooling water into an in-containment refueling water tank, a sump or the like, and respraying the cooling water to reduce the pressure, temperature and concentration of radioactive material in the containment for a long period of time. The active containment spray system may perform a long-term spray function, but has a characteristic in which a power system for driving the pump must be available. The passive containment spray system (CANDU, Canada, etc.) method has a cooling water storage tank at an upper section of the containment, and performs a function of spraying a large amount of cooling water during an accident to reduce the pressure, temperature and the concentration of radioactive material inside the containment. The passive containment spray system has a characteristic in which the system cannot be operated for more than a predetermined period of time due to the limitation of the storage capacity of the cooling water. Therefore, in order to use the passive containment spray system for a long period of time, the cooling water storage tank should be replenished periodically using a pump. This means that a pump and a power system for driving the pump must be available for a long-term operation of the passive containment spray system. The suppression tank (commercial BWR, CAREM: Argentina, IRIS: U.S. Westinghouse Corporation, etc.) method performs a function of introducing steam discharged into the containment to the suppression tank using a pressure difference inside the containment and the suppression tank, and condensing the steam to reduce the pressure, temperature and the concentration of the radioactive material inside the containment. The suppression tank method has a characteristic of operating only until the pressure inside the containment is higher than the pressure inside the suppression tank. The passive containment cooling system method performs a function of installing a heat exchanger and a cooling water tank inside or outside the containment, and condensing steam inside the containment using the heat exchanger to reduce the pressure, temperature and the concentration of the radioactive material inside the containment. The passive containment cooling system method uses natural circulation inside the containment, and thus has a characteristic in which a reduction capacity of the pressure, temperature and radioactive material are relatively reduced as compared with the active spray system. In addition, as a part of the passive containment there is a method in which a steel containment vessel is applied, an outer wall is cooled (spray, air), steam inside the containment vessel is condensed on an inner wall of the containment vessel to perform a function of reducing the pressure, temperature and the concentration of radioactive material (AP1000: Westinghouse Inc., USA), and the like. Similarly to the passive containment cooling system method, this method uses natural circulation inside the containment, and thus has a characteristic in which a reduction capacity of the pressure, temperature and radioactive material are relatively reduced as compared with the active spray system. Particularly, for a cooling water storage method, a refueling water tank may be installed in the containment vessel to use refueling water in the in-containment refueling water storage tank (IRWST) as a coolant water supply source. The in-containment refueling water storage tank (IRWST) may perform the role of storing refueling water, providing safe injection and a cooling water source of the containment vessel spray system, and providing a heat sink for condensing the condensation of steam discharged from the pressurizer. In detail, cooling water required for the refueling water tank may be supplied by a unique function of the in-containment refueling water storage tank (IRWST) in the storage tank during refueling. In addition, the in-containment refueling water storage tank (IRWST) may accommodate materials discharged during a nuclear accident to prevent contamination inside the containment. Most of the systems described above have excellent performance in reducing the pressure and temperature inside the containment. However, the system described above has a limitation in reducing radioactive material, iodine, which can diffuse to the external environment during a nuclear accident. Iodine generated during a nuclear accident may dissolve in most cases when in contact with water (solubility 0.029 g/100 g (20° C.)). Therefore, the most excellent performance for reducing the concentration of radioactive material inside the containment among these safety systems related to the containment is provided by an active containment spray system (a method employed by Korea commercial reactors) which sprays a large amount of cooling water using an active pump and recirculates the cooling water for a long period of time. However, the active safety system has a problem in which emergency AC power must be supplied in order to operate an active device such as a pump during a nuclear accident, and the active safety system does not operate when the emergency AC power is not supplied. From this point of view, there is a growing demand for a passive safety system in which the safety system operates without supplying emergency AC power. When the passive safety system is applied to a nuclear power plant during a nuclear accident, the safety of the nuclear power plant may be significantly increased compared to the case of applying the active safety system. However, when the passive safety system method is employed as a safety system of the containment, the performance of cooling the containment is insufficient as compared to the active safety system method, and thus the concentration of radioactive material inside the containment may be relatively high. Accordingly, when the passive safety system method is employed as a safety system for the containment, an exclusion area boundary (EAB) or a low population zone (LPL), which restricts the residence of the public for the safety of the general public during an accident by assuming a nuclear accident, will be set to a nuclear power plant. In particular, in setting the exclusion area boundary (EAB), it is common to set a larger restricted area boundary (EAB) compared to that of the active safety system. The expansion of the restricted area boundary (EAB) may result in a problem that greatly increases the construction cost of a nuclear power plant. Accordingly, the present disclosure presents a radioactive material reduction facility of effectively collecting radioactive material and being effectively used as a heat sink in the in-containment refueling water storage tank (IRWST) to prevent the pressure of the nuclear reactor and the concentration of radioactive material from being increased during a nuclear accident, and a nuclear power plant having the same. An object of the present disclosure is to propose a radioactive material reduction facility which contributes to the safety improvement of a nuclear power plant capable of reducing the concentration of radioactive material discharged during an accident of leaking reactor coolant from the nuclear power plant. Another object of the present disclosure is to propose a radioactive material reduction facility for providing a heat sink for condensation of steam containing radioactive material according to an accident progression during an accident of leaking reactor coolant from the nuclear power plant. Still another object of the present disclosure is to propose a radioactive material reduction facility capable of solving a problem of expanding an exclusion area boundary of a nuclear power plant, and the nuclear power plant having the same. The present disclosure relates to a radioactive material reduction facility, including a containment, a boundary section provided inside the compartment to partition an inner space of the containment into a first space for accommodating a reactor coolant system and a second space formed between the first space and the containment, and surround the reactor coolant system to prevent radioactive material discharged from the reactor coolant system or a line connected to the reactor coolant system inside the first space from being directly discharged into the second space during an accident, an in-containment refueling water storage tank (IRWST) installed between the first space and the second space and formed to accommodate refueling water, and a first discharge line formed to guide the flow of steam and radioactive material formed in the first space inside the boundary section into the in-containment refueling water storage tank. Specifically, in the radioactive material reduction facility according to the present disclosure, the in-containment refueling water storage tank may include a plurality of pools separated from each other, and the plurality of pools may include at least a first pool and a second pool, and the steam and the radioactive material may be discharged to the second pool through the first pool during an accident. According to an embodiment, the boundary section may include a partition wall formed to surround the reactor coolant system, and a cover formed to cover an upper section of the reactor coolant system. According to an embodiment, the radioactive material reduction facility may include a refueling cavity formed to surround the cover at the boundary section. According to an embodiment, the upper section of the reactor coolant system and the cover may be formed to be detachable. According to an embodiment, the radioactive material reduction facility may further include a sealing device formed to be open and closed between the partition wall and the reactor coolant system. According to an embodiment, the radioactive material reduction facility may further include a connection line formed between the in-containment refueling water storage tank and the reactor coolant system to flow refueling water. According to an embodiment, refueling water may be supplied based on a driving force of a pump for the injection of the connection line. According to an embodiment, the radioactive material reduction facility may further include a discharge section provided at an upper section of the in-containment refueling water storage tank to prevent an overpressure of the in-containment refueling water storage tank and discharge non-condensable gas or steam into the second space. According to an embodiment, the capacity of the first pool may be larger than that of the second pool, and the capacity of the second pool may be 1/20 to 1/10 of the capacity of the first pool. According to an embodiment, the radioactive material reduction facility may include a second discharge line formed between the first pool and the second pool to flow steam and radioactive material. According to an embodiment, the second discharge line may be formed to discharge steam and radioactive material at a pressure lower than the pressure at which the steam and the radioactive material are discharged from the first discharge line. According to an embodiment, the first discharge line may further include a check valve formed to move steam and radioactive material formed in the first space to the in-containment refueling water storage tank. According to an embodiment, refueling water in the second pool may be maintained at a pH of a preset value or more to prevent the volatilization of radioactive material during a normal operation and during an accident. According to an embodiment, a pH of the refueling water of the second pool may be higher than that of the refueling water of the first pool during an accident. According to an embodiment, the second pool may further include an additive injection facility for injecting an additive into the second pool to allow a pH value of the refueling water to be maintained above the preset value. According to an embodiment, the additive injection facility may be formed not to inject an additive during a normal operation and formed to inject the additive during an accident. Furthermore, the present disclosure relates to a nuclear power plant, including a containment, a boundary section provided inside the compartment to partition an inner space of the containment into a first space for accommodating a reactor coolant system and a second space formed between the first space and the containment, and surround the reactor coolant system to prevent radioactive material discharged from the reactor coolant system or a line connected to the reactor coolant system inside the first space from being directly discharged into the second space during an accident, an in-containment refueling water storage tank (IRWST) installed between the first space and the second space and formed to accommodate refueling water, and a first discharge line formed to guide the flow of steam and radioactive material formed in the first space inside the boundary section into the in-containment refueling water storage tank. Specifically, in the nuclear power plant according to the present disclosure, the in-containment refueling water storage tank may include a plurality of pools separated from each other, and the plurality of pools may include at least a first pool and a second pool, and the steam and the radioactive material may be discharged to the second pool through the first pool during an accident. When a reactor coolant leaking accident occurs, the radioactive material reduction facility according to the present disclosure may jet radioactive material and steam contained in the atmosphere (air and steam) in a first space to the refueling water of the in-containment refueling water storage tank (IRWST) for collection, and condensing the radioactive material in a second pool to reduce the concentration of the radioactive material inside the containment, thereby improving the safety of the nuclear power plant. Furthermore, when an accident of leaking reactor coolant occurs, the radioactive material reduction facility according to the present disclosure may operate effectively to a heat sink through a plurality of pools connected to each other to discharge non-condensable gas, steam and radioactive material to an inside of the containment through a discharge section of the in-containment refueling water storage tank (IRWST), thereby preventing the pressure of a second space from rapidly increasing. In addition, when an accident of leaking reactor coolant occurs, the radioactive material reduction facility according to the present disclosure may effectively collect radioactive material to solve a problem of expanding the exclusion area boundary so as to improve the safety of the nuclear power plant, thereby reducing the nuclear construction cost. Hereinafter, preferred embodiments of the present disclosure will be described in detail with reference to the accompanying drawings, and the same or similar elements are designated with the same numeral references regardless of the numerals in the drawings and their redundant description will be omitted. In describing the embodiments disclosed herein, moreover, the detailed description will be omitted when specific description for publicly known technologies to which the invention pertains is judged to obscure the gist of the present invention. Also, it should be understood that the accompanying drawings are merely illustrated to easily explain the concept of the invention, and therefore, they should not be construed to limit the technological concept disclosed herein by the accompanying drawings, and the concept of the present disclosure should be construed as being extended to all modifications, equivalents, and substitutes included in the concept and technological scope of the invention. It will be understood that although the terms first, second, etc. may be used herein to describe various elements, these elements should not be limited by these terms. The terms are used merely for the purpose to distinguish an element from the other element. Incidentally, unless clearly used otherwise, expressions in the singular number include a plural meaning. Terms “include” or “has” used herein should be understood that they are intended to indicate an existence of several components or several steps, disclosed in the specification, and it may also be understood that part of the components or steps may not be included or additional components or steps may further be included. FIG. 1A is a conceptual view of a radioactive material reduction facility 100 associated with an embodiment of the present disclosure, and a nuclear power plant 10 having the same. The containment 12 is installed at an outside of the reactor coolant system 11 to prevent the leakage of the radioactive material. In the present disclosure, the containment 12 collectively refers to a containment building, a reactor building, a containment vessel, a safeguard vessel, and the like in the nuclear power plant 10. The boundary section 110 is provided inside the containment 12 to partition the containment 12 into a first space 110a for accommodating a reactor coolant system 11 and a second space 110b formed between the first space 110a and the containment 12. Furthermore, the boundary section 110 surrounds the reactor coolant system 11 to prevent radioactive material from being leaked from a connecting line (not shown) installed in the first space 110a into the second space 110b. The atmosphere in the first space 110a inside the boundary section 110 includes air existing in the first space 110a during a normal operation of the nuclear power plant. On the other hand, the atmosphere may include steam or radioactive material being discharged during a nuclear accident such as a steam line break accident or a loss-of-coolant accident. The boundary section 110 surrounds the reactor coolant system 11 to prevent the radioactive material from being leaked through a path other than the connecting line (not shown) installed in the first space 110a, thereby forming a sealing structure. Therefore, the boundary section 110 may be closed at the time of an accident to maintain a sealing structure. The boundary section 110 is formed at a design pressure capable of withstanding a pressure of the steam of the vaporized coolant during an accident. At least a part of the boundary section 110 may be formed by a concrete structure inside the containment 12 and a coating member (not shown) provided on the concrete structure. The boundary section 110 may include a partition wall 111 and a cover 112. The cover 112 is formed to cover an upper section of the reactor coolant system 11. A bottom surface (or double bottom surface) of the partition 111, the cover 112 and the containment 12 may form a sealing structure around the reactor coolant system 11. In addition, a sealing device 140 formed to be open and closed between the partition wall 111 and the reactor coolant system 11. The sealing device 140 is open during a normal operation of the nuclear power plant to partition the first space 110a between the reactor coolant system 11 and the cover 112. The upper section 11e of the reactor coolant system and the cover 112 are formed to be detached from the reactor coolant system 11 and the partition wall 111, respectively. On the other hand, during a normal operation of the nuclear power plant, the upper section 11e of the reactor coolant system and the cover 112 are attached to the reactor coolant system 11 and the partition wall 111, respectively, to maintain air-tightness. On the other hand, during a normal operation of the nuclear power plant, a fission reaction occurs in the core 11a to generate heat. Heat generated from the core 11a may be transferred to coolant inside the reactor coolant system 11 to produce steam in the steam generator 11c. The steam generator 11c may be a pressurized water reactor. Moreover, steam produced by the steam generator 11c may be steam that is phase-changed by receiving water from a feedwater system (not shown) through a main feedwater line (not shown) and an isolation valve (not shown) connected thereto. The steam produced by the steam generator 11c is supplied to a large turbine (not shown) and a large generator (not shown) to produce electric power while the fluid energy of the steam is converted into electric energy through mechanical energy. However, although the pressurized water reactor is illustrated in the present disclosure, the technology of the present disclosure is not limited to the pressurized water reactor. In addition, a reactor coolant pump 11d may circulate coolant 11b that fills an inside of the reactor coolant system 11. According to an embodiment, the refueling cavity 130 surrounds the cover 112 and is formed so as to protrude from the partition wall 111 to accommodate refueling water during the refueling operation. On the other hand, the sealing device 140 is formed to be closed during the refueling operation. During the refueling operation, the upper section Ile of the reactor coolant system and the cover 112 are separated from the reactor coolant system 11 and the partition 111, respectively. It will be described in detail in the following description of FIG. 1B. The radioactive material reduction facility 100 is installed inside the containment 12 and includes an in-containment refueling water storage tank 120 formed to accommodate refueling water between the first space 110a and the second space 110b. The in-containment refueling water storage tank 120 includes a first discharge line 121, a first pool 122, a second discharge line 123, a second pool 124, and a discharge section 125. The in-containment refueling water storage tank 120 is provided with a plurality of pools separated from each other to accommodate refueling water. The pool of the in-containment refueling water storage tank 120 includes at least a first pool 122 and a second pool 124. During an accident, steam and radioactive material are introduced into the in-containment refueling water storage tank 120. The first discharge line 121 is formed to introduce the steam of the vaporized coolant and radioactive material contained therein from a reactor coolant system 11 installed inside the containment 12 or a line (not shown) connected to the reactor coolant system 11 to flow into the in-containment refueling water storage tank 120. In detail, the steam of the vaporized coolant and the radioactive material contained therein are discharged due to a difference between an internal pressure of the boundary section 110 including the reactor coolant system 11 and a pressure of the in-containment refueling water storage tank 120. As a result, the steam of the vaporized coolant and the radioactive material contained therein discharged through the first discharge line 121 are introduced into the first pool 122 to reduce the pressure while absorbing the heat and reducing the volume through the refueling water accommodated in the first pool 122. Then, the steam of the vaporized coolant and the radioactive material contained therein discharged through the refueling water contained in the first pool 122 form a flow due to the pressure difference. Accordingly, it may be possible to collect radioactive material and reduce the volume of the steam of the vaporized coolant when introduced into the second pool 124 in which refueling water is accommodated through the second discharging line 123 between the first pool 122 and the second pool 124. In other words, the radioactive material and the steam pass through the first pool 122 and the second pool 124 connected to each other when an accident of leaking reactor coolant occurs, the radioactive material reduction facility 100 according to the present disclosure may effectively operates as a heat sink. Accordingly, it may be possible to prevent the pressure of the second space 110b from being drastically increased due to the non-condensable gas, the steam, and the radioactive material discharged during a nuclear accident. According to an embodiment, in order to effectively reduce the pressure of the steam of the vaporized coolant flowing into the first pool 122 through the first discharge line 121, the refueling water accommodated in the first pool 122 and the steam of the vaporized coolant should be in contact therewith for more than a predetermined period of time. Therefore, an end of the first discharge line 121 is preferably installed at the bottom of the first pool 122, and the steam of the vaporized coolant and the radioactive material are jetted through the sparger 121a. The steam discharged through the sparger 121a is condensed in the refueling water of the first pool 122, and the water soluble radioactive material is dissolved and collected in the refueling water, and the non-condensable gas or part of the non-condensed steam rises while being cooled. In particular, the refueling water of the first pool 122 may be formed to suppress the reactivity of the core 11a when fuel including boric acid is reloaded. As a result, the refueling water of the first pool 122 is maintained at a pH of 5 to 5.5. In addition, the pressure of the boundary portion 110 may be reduced while most of the vaporized coolant during an accident is compressed through the refueling water accommodated in the first pool 122. The refueling water accommodated in the first pool 122 may have a height of above 6 m to bring the steam of the vaporized coolant into contact with the steam for more than a predetermined period of time. When the height of the refueling water accommodated in the first pool 122 is less than 6 m, it is difficult to sufficiently perform the role of the heat sink for effectively condensing the steam of the vaporized coolant discharged during an accident. In addition, the first pool 122 may be formed to have a larger capacity than the second pool 124. Specifically, the capacity of the second pool 124 is preferably formed to be 1/20 to 1/10 of the capacity of the first pool 122. When the capacity of the second pool 124 is less than 1/20 of the capacity of the first pool 122, the radioactive material in the second pool 124, which will be described later, cannot be collected efficiently. On the other hand, when the capacity of the second pool 124 is greater than 1/10 of the capacity of the first pool 122, the capacity of the refueling water accommodated in the first pool 122 is reduced, thereby resulting in difficulty in compressing the steam discharged during an accident. In other words, when the capacity of the second pool 124 is greater than 1/10 of the capacity of the first pool 122, the capacity of the refueling water accommodated in the first pool 122 is reduced, thereby resulting in difficulty in reducing the pressure of the steam discharged during an accident. According to an embodiment, the first pool 122 may be filled with about 3,500 t of refueling water, and the second pool 124 may be filled with about 200 t of refueling water. Accordingly, as described above, it may be possible to sufficiently perform the role of the heat sink for efficiently condensing the steam discharged during an accident through the refueling water accommodated in the first pool 122, and the radioactive material that has not been sufficiently collected in the first pool 122 may be collected through the second pool 124. Specifically, the second discharge line 123 is formed to discharge steam and radioactive material at a pressure lower than the pressure at which the steam and the radioactive substance are discharged from the first discharge line 121. As a result, the steam and the radioactive material whose pressure has been reduced through the refueling water accommodated in the first pool 122 are discharged through the second discharge line 123 and the sparger 123a into the refueling water of the second pool 124. Specifically, in order to allow steam and radioactive material to be discharged to the second discharge line 123 at a pressure lower than the pressure at which the steam and the radioactive substance are discharged from the first discharge line 121, the second discharge line 123 may be designed to have a smaller head than the first discharge line 121. Furthermore, the first discharge line 121 may further include a check valve 121b formed to move the steam and the radioactive material formed in the first space 110a to the in-containment refueling water storage tank 120. Accordingly, even when the pressure of the in-containment refueling water storage tank 120 increases due to superheated steam, it may be possible to prevent refueling water accommodated in the in-containment refueling water storage tank 120 from flowing backward to the first space 110a in which the reactor coolant system 11 is accommodated. On the other hand, since the pressure in the first space 110a is not higher than the pressure in the first pool 122, the flow of air to the check valve 121b and the first discharge line 121 is not formed. According to an embodiment, the radioactive material reduction facility 100 may be designed to collect water-soluble radioactive material in the refueling water of the second pool 124, thereby preventing the radioactive material from being leaked directly to the external environment. In other words, the radioactive material may be condensed in the second pool 124 to reduce the concentration of the radioactive material inside the containment 12, thereby improving the safety of the nuclear power plant 10. The radioactive material discharged during a nuclear accident may include xenon (Xe), krypton (Kr), and iodine 131I. In particular, in the case of 131I, it discharges energy of 0.971 MeV including β-ray in the nucleus and decays to 131Xe. Therefore, compared with radioactive xenon (Xe) and krypton (Kr), 131I accumulates in a thyroid gland of a human body to cause diseases such as cancer, and is known to be a dangerous radioactive material during a nuclear accident. Radioactive iodine generated during a nuclear accident may dissolve in most cases when in contact with water (solubility 0.029 g/100 g (20° C.)). However, the radioactive iodine dissolved in the cooling water exists in the form of an anion, and the amount of the radioactive iodine may be greatly increased when the pH of the cooling water in which the radioactive iodine is dissolved is low. This is because the amount of radioactive iodine converted into the volatile elementary iodine (I2) in the cooling water of pH 7 or less is greatly increased. In other words, most of the radioactive iodine cannot be collected in the refueling water of pH of 5 to 5.5 accommodated in the first pool 122, and discharged to an upper section of the first pool 122 through the refueling water accommodated in the first pool 122. Accordingly, the first pool 122 and the second pool 124 are formed to accommodate the refueling water having different pHs from each other, and the pH of the refueling water of the second pool 124 may be higher than that of the refueling water of the first pool 122. Moreover, the refueling water of the second pool 124 may be formed to accommodate the refueling water having a pH of a preset value or more so as to prevent the volatilization of the radioactive material (in particular, iodine). Specifically, the pH of the preset value may be pH 7, and preferably pH 8 to 9. Meanwhile, the radioactive material reduction facility 100 may further include an additive injection facility 160. According to an embodiment, the additive is introduced into the second pool 124 during a nuclear accident, and the refueling water accommodated in the first pool 122 and the second pool 124 are formed to have different pHs. Specifically, when the refueling water accommodated in the second pool 124 is refueling water having a pH of 5 to 5.5 including boric acid, such as the refueling water accommodated in the first pool 122 described above, the additive injection facility 160 supplies an additive to the refueling water accommodated in the second pool 124 that maintains the pH of the refueling water accommodated in the second pool 124 at a preset value or more (generally, a pH of 7 or more) so as to prevent the volatilization of the radioactive material dissolved in the second pool 124. According to an embodiment, the additive injection facility 160 is formed to accommodate the additive in the additive accommodation section 161. During a normal operation of the nuclear power plane, the additive does not flow into the second pool 124 while the valve 162 is closed. However, during a nuclear accident, the valve 162 is open to inject the additive to the refueling water accommodated in the second pool 124 through the connection line 163. For the additive, trisodium phosphate (Na3PO4) may be used. Trisodium phosphate controls the pH of the cooling water to prevent corrosion inside the containment 12 and re-volatilization of the radionuclide. However, in the present disclosure, the type of the additive is not necessarily limited thereto. Moreover, a filter or an adsorbent (not shown) may be provided in the discharge section 125 to capture the radioactive material to be discharged into the second space 110b together with the steam during a nuclear accident. The filter or adsorbent is made to allow steam or air to pass therethrough and to capture the radioactive material. According to an embodiment, the filter may use a high efficiency particle filter (HEPA filter). A gaseous form of radioactive material contained in the steam or atmosphere is removed while passing through the filter. For example, when the radioactive material is iodine, the iodine is converted to iodic silver in combination with silver nitrate while passing through the filter and removed from the steam or atmosphere. Furthermore, activated carbon may be used for the adsorbent. Activated carbon is used for an adsorbent because it has a large internal adhesion area due to its porous structure. Iodine organic compounds are converted into a quaternary ammonium salt form in combination with materials impregnated in activated carbon and adsorbed on activated carbon. In addition, a molecular form of iodine binds to activated carbon through chemical adsorption. The filter and the adsorbent may be disposed together, or only one of the filter and the adsorbent may be disposed. However, the above-described filter and adsorbent are described only by way of example, and the types of the filter and the adsorbent in the present disclosure are not necessarily limited to those described above. FIG. 1B is a conceptual view of a radioactive material reduction facility 100 associated with an embodiment of the present disclosure, and a refueling operation of a nuclear power plant 10 having the same. During the refueling operation of the nuclear power plant 10, the sealing device 140 may be closed when the refueling operation is carried out to partition the first space 110a′ below the sealing device 140. Moreover, the upper section 11e of the reactor coolant system and the cover 112 are separated from the reactor coolant system 11 and the partition 111, respectively. In other words, a lower section of the refueling cavity 130 is closed by the sealing device 140. During the refueling operation, refueling water is supplied to the reactor coolant system 11 through the connection line 150 formed to flow the refueling water between the in-containment refueling water storage tank 120 and the reactor coolant system 11. The refueling water supplied to the connection line 150 fills the reactor coolant system 11 and fills the refueling cavity 130. In other words, the refueling water supplied to the reactor coolant system 11 is also supplied to and accommodated in the refueling cavity 130 to carry out the refueling operation safely. In detail, during the recharging operation, the injection of the refueling water through the connection line 150 is supplied based on a driving fore of the pump 151. Furthermore, according to another embodiment electric power production section 230 described below, the same or similar reference numerals are designated to the same or similar configurations to the foregoing example, and the description thereof will be substituted by the earlier description. FIG. 2 is a conceptual view of a radioactive material reduction facility 200 associated with another embodiment of the present disclosure. Referring to FIG. 2, the second pool 224 of the radioactive material reduction facility 200 may be disposed at an upper section of the first pool 222. As a result, during an accident, steam and radioactive material are introduced into the in-containment refueling water storage tank 220. When an accident occurs, the first discharge line 221 allows the steam of coolant vaporized from the reactor coolant system 11 installed inside the containment 12 or a line (not shown) connected to the reactor coolant system 11 or the radioactive material contained therein to be introduced into the first pool 222 inside the containment while forming a flow due to a pressure difference to lose the heat and reduce the volume so as to reduce the pressure while passing through the refueling water accommodated in the first pool 222. Then, the steam of the vaporized coolant and the radioactive material contained therein discharged through the refueling water contained in the first pool 222 form a flow due to the pressure difference. Accordingly, it may be possible to collect radioactive material and reduce the volume of the steam of the vaporized coolant when introduced into the second pool 224 in which refueling water is accommodated through the second discharging line 223 between the first pool 222 and the second pool 224. According to an embodiment, the refueling water accommodated in the second pool 224 may be formed to accommodate the refueling water having a pH of a preset value or more so as to prevent the volatilization of the radioactive material (in particular, iodine). Specifically, the pH of the preset value may be pH 7, and preferably pH 8 to 9. Moreover, the radioactive material reduction facility 200 may further include an additive injection facility (not shown). According to an embodiment, it may also be possible to have an embodiment in which the additive is introduced into the second pool 224 during a nuclear accident, and the refueling water accommodated in the first pool 222 and the second pool 224 are formed to have different pHs. It is obvious to those skilled in the art that the present disclosure can be embodied in other specific forms without departing from the concept and essential characteristics thereof. In addition, the detailed description thereof should not be construed as restrictive in all aspects but considered as illustrative. The scope of the invention should be determined by reasonable interpretation of the appended claims and all changes that come within the equivalent scope of the invention are included in the scope of the invention. |
|
046413359 | description | DESCRIPTION OF THE PREFERRED EMBODIMENTS FIG. 1 illustrates a primary-beam collimator which permits the restriction of three beam pyramids originating from three focal points: F1, F2 and FN. Focal points F1 and F2 serve to create pairs of stereo images, while focal point FN is used to produce regular radiographs. Beam restriction in planes parallel to the stero base (i.e., line F1-F2) is achieved as in the case of regular primary-base shutters, with the aid of first-stage and second-stage pairs of shutter leaves, 1a, 1b. Restriction in planes orthogonal to the stereo base is achieved during stereo operation (i.e., alternative activation of focal points F1, F2) with the aid of first-stage shutter leaves 2a (restricting the beam from F1), and 2b (restricting the beam from F2), and the second-stage shutter leaves 3a and 3b. Shutter leaves 2a, 2b, and 3a, 3b can be adjusted in a plane perpendicular to the central-beam axis of the restricted pyramid-shaped beam, in which the stereo base is located. Inner positions of leaves 2a, 2b have been represented by dash-dot lines. Internal restriction of the beam pyramids is achieved by means of the swiveling shutter leaves 4a (for point F1) and 4b (for point F2). Shutter leaves 4a, 4b swivel about axes 5a, 5b, and can also be adjusted in a direction parallel to the stereo base on guides 14 (FIG. 3). Shutter leaves 2a, 2b, 3a, 3b, and the swiveling shutter leaves 4a, 4b, are operated by a stepping motor (not shown). Their movements are coordinated by a microcomputer. In order to obtain pairs of stereo radiographs, shutter leaves 1a, 1b, 2a, 2b, 3a, 3b, and 4a, 4b are positioned as represented by solid lines, resulting in the shaping of two beam pyramids (I and II). When shifting from stereo to regular operation, the microcomputer coordinates the movements of the shutter to new positions. Shutter leaves 2a, 3a, and swiveling shutter leaf 4a which, during stereo operation, restrict beam pyramid I at the top and the bottom, as shown in FIG. 1, now become the top restriction of beam pyramid III (with the vertex in FN). The same is valid for the bottom restriction, which is obtained with the aid of shutter leaves 2b, 3b, and 4b. According to FIG. 3, the swiveling shutters 4a, 4b are constructed so as to allow free movement to shutter-leaf pairs 1a, 1b and filter frames 6a, 6b, 6c, 6d in a cross-section adjacent to axes 5a, 5b. During the stereo operation, restriction is achieved by external surfaces 7a, 7b (FIG. 2) of swiveling shutters 4a, 4b, while during regular operation restriction is performed by the internal surfaces 8a, 8b. The swivel movement of swiveling shutters 4a, 4b takes place over a guide link 9 (FIG. 2). In the space between the swiveling shutters 4a, 4b and shutter-leaf pair 1a, 1b a sighting device 10a, 10b, 10c may be provided, consisting of a light bulb and two mirrors (FIG. 1). For the operation of an image-intensifier tube with a circular entrance filter, it is possible to add an octagonal diaphragm 11, which is only schematically shown in the figures. In the space between the swiveling shutters 4a, 4b an additional spring-operated swiveling shutter 12 may be introduced, as shown in FIGS. 2 and 3. During stereo operation with slight beam restriction, this swiveling shutter 12 closes the gap between the swiveling shutters 4a and 4b, cutting off scatter radiation. During regular operation, swiveling shutter 12 is kept open by shutter leaf 2a, now positioned along shutter leaf 4a, with the aid of a linkage rod 13. An essential feature of the primary-beam collimator as shown in FIGS. 1 to 3 is the following: The shutter leaves 1a, 1b, 2a, 2b, 3a, 3b, and 4a, 4b are constructed and can be adjusted so that beam pyramids I and II, from focal points F1 and F2, respectively, can be individually controlled. For the restriction by planes perpendicular to the stereo base F1-F2, two internal shutter leaves 4a, 4b are provided, adjustable between two external shutter leaves 2a, 2b. The internal shutter leaves 4a, 4b are constructed so that, when brought to their external position, they permit the free emission of a central beam-pyramid III from a central focal point FN and, in conjunction with the external shutter leaves 2a, 2b, they close the external openings in the path of beam pyramids I, II. The above-described primary-beam collimator can be used, as needed, for either regular or stereo shutter operation. There has thus been shown and described a beam collimator for x-ray apparatus which fulfills all the objects and advantages sought therefor. Many changes, modifications, variations and other uses and applications of the subject invention will, however, become apparent to those skilled in the art after considering this specification and the accompanying drawing which discloses the preferred embodiment thereof. All such changes, modifications, variations and other uses and applications which do not depart from the spirit and scope of the invention are deemed to be covered by the invention which is limited only by the claims which follow. |
059329300 | abstract | A process for fabricating weapons-grade plutonium into mixed-oxide fuel for use in a nuclear reactor. The plutonium is converted into plutonia powder at a site remote from the fuel fabrication facility and then the plutonia powder is pressed and fired into high-density plutonia pellets for transport to the fuel fabrication facility. The high-density plutonia pellets serve as a diversion-resistant and environmentally sound method of shipping the plutonia for use as a feed material. Comminution methods are employed to reduce the plutonia pellets to a fine powder for blending with urania feed to produce mixed oxide acceptable for reactor operation. |
abstract | A remote sensing device for detecting materials of varying atomic numbers and systems and methods relating thereto. A system for identifying a material includes a photon beam flux monitor for resolving a high-energy beam. A method for identifying a material includes casting an incident photon beam on the material and detecting an emerging photon beam with an array of fission-fragment detectors, a first set of scintillator paddles, and a second set of scintillator paddles. |
|
abstract | A particle beam irradiation apparatus includes a synchrotron, two scanning electromagnets, an beam delivery apparatus for outputting an ion beam extracted from the synchrotron, and an accelerator and transport system controller, and a scanning controller. These controllers stop the output of the ion beam from the beam delivery apparatus; in a state where the output of the ion beam is stopped, change the irradiation position of the ion beam by controlling the scanning electromagnets; and after this change, control the scanning electromagnets to start the output of the ion beam from the beam delivery apparatus and to perform irradiations of the ion beam to at least one irradiation position a plurality of times based on treatment planning information. |
|
summary | ||
050248065 | claims | 1. In a fuel assembly for a nuclear reactor including a plurality of nuclear fuel rods, at last one grid supporting said fuel rods in an organized array, and at least one guide thimble supporting said grid, a debris filter bottom nozzle spaced below said grid and said fuel rods, supporting said guide thimble and adapted to allow flow of liquid coolant into said fuel assembly, said debris filter bottom nozzle comprising: (a) an enclosure defining a chamber having opposite open ends for coolant flow therethrough in an axial direction from one of said ends to the other of said ends; and (b) an array of elongated hollow tubes disposed across said chamber of said enclosure in a consolidated side-by-side relation to one another and in abutting contact with one another and rigidly interconnected to said enclosure and to one another to form a unitary structure, said hollow tubes extending substantially parallel to one another and axially in the direction of coolant flow through said chamber; (c) said array of elongated hollow tubes including a first plurality of tubes of a substantially constant first diameter size and a second plurality of tubes of a substantially constant second diameter size which is smaller than the first diameter size of said first plurality of tubes, said first plurality of tubes being interspersed with said second plurality of tubes in an ordered pattern, said first and second pluralities of elongated hollow tubes being substantially arranged in interconnected structural units wherein each unit includes one tube of said second plurality of tubes in abutting contact with at lest three tubes of said first plurality of tubes, said at least three tubes of said first plurality of tubes being circumferentially arranged around said one tube of said second plurality of tubes and in abutting contact with one another, said first and second pluralities of hollow tubes having hollow tubular cross-sectional configurations defining passages for coolant flow through said chamber, said first diameter size tubes of said first plurality of tubes having passages being of a size smaller than a substantial amount of damage-inducing debris contained in the coolant flow which thereby prevents travel of such debris with the coolant flow through the passages. (a) an enclosure defining a chamber having opposite open ends for coolant flow therethrough in an axial direction from one of said ends to the other said ends; and (b) an upper transverse nozzle structure composed of an array of elongated hollow tubes disposed across said chamber of said enclosure in a consolidated side-by-side relation to one another and in abutting contact with one another and rigidly interconnected to said enclosure, said hollow tubes extending substantially parallel to one another and axially in the direction of coolant flow through said chamber, said hollow tubes being shorter in axial length than the height of said enclosure; (c) said array of elongated hollow tubes including a first plurality of tubes, a second plurality of tubes and a third plurality of tubes, said first plurality of tubes having a substantially constant first diameter size and said second plurality of tubes having a substantially constant second diameter size which is smaller than the first diameter size of said first plurality of tubes, said first plurality of tubes being interspersed with said second plurality of tubes in an ordered pattern, said first and second pluralities of elongated hollow tubes being substantially arranged in interconnected structural units wherein each unit includes one tube of said second plurality of tubes in abutting contact with at least three tubes of said first plurality of tubes, said at least three tubes of said first plurality of tubes being circumferentially arranged around said one tube of said second plurality of tubes and in abutting contact with one another, said first and second pluralities of hollow tubes having hollow tubular cross-sectional configurations defining passages for coolant flow through said chamber, said first diameter size tubes of said first plurality of tubes having passages being of a size smaller than a substantial amount of damage-inducing debris contained in the coolant flow which thereby prevents travel of such debris with the coolant flow through the passages, said third plurality of hollow tubes having tubular cross-sectional configurations defining bores for receiving lower ends of said guide thimbles for attachment to said bottom nozzle. 2. The bottom nozzle as recited in claim 1, wherein at least one of said hollow tubes has a bore for receiving a lower end of said one guide thimble for attachment to said bottom nozzle. 3. The bottom nozzle as recited in claim 1, wherein said hollow tubes are shorter in axial length than the height of said enclosure. 4. The bottom nozzle as recited in claim 1, wherein said enclosure includes a plurality of interconnected upstanding side walls which define said coolant flow chamber therethrough. 5. The bottom nozzle as recited in claim 4, wherein said enclosure also includes a plurality of legs connected to said side walls for supporting said bottom nozzle in the reactor. 6. The bottom nozzle as recited in claim 5, wherein said interconnected side walls define a plurality of corners in said enclosure and each of said support legs is disposed at one of said corners of said enclosure. 7. In a liquid cooled nuclear reactor having a plurality of fuel assemblies, each fuel assembly including a plurality of nuclear fuel rods, a plurality of grids axially disposed along and supporting said fuel rods in an organized array, and a plurality of guide thimbles supporting said grid, a debris filter bottom nozzle spaced below a lowermost one of said girds and said fuel rods, supporting said guide thimbles and adapted to allow flow of liquid coolant into said fuel assembly, said debris filter bottom nozzle comprising: 8. The bottom nozzle as recited in claim 7, wherein said enclosure includes a plurality of interconnected upstanding side walls which define said coolant flow chamber therethrough. 9. The bottom nozzle as recited in claim 8, wherein said enclosure also includes a plurality of legs connected to said side walls for supporting said bottom nozzle in the reactor. 10. The bottom nozzle as recited in claim 9, wherein said interconnected side walls define a plurality of corners in said enclosure and each of said support legs is disposed at one of said corners of said enclosure. |
description | This application claims the priority, under 35 U.S.C. §119, of German application DE 10 2008 013 213.6-54, filed Mar. 7, 2008; the prior application is herewith incorporated by reference in its entirety. The invention relates to a method for catalytic recombination of hydrogen, which is carried in a gas flow, with oxygen, in which the gas flow is passed through a reaction zone with a number of catalytic converter elements, with steam being added to the gas flow before it enters the reaction zone. It also relates to a recombination system for catalytic recombination of hydrogen, which is carried in a gas flow, with oxygen, in which an inlet line, which is provided in order to supply the gas flow into a reaction zone which is formed by a number of catalytic converter elements, is connected to a steam supply line, in order to add steam as required. In many technical installation, the operation, malfunctions or defects can lead to a release of combustible gases such as hydrogen, tritium or hydrocarbon compounds. For example, during operation of nuclear installations in nuclear power stations, particularly during power operation of the light-water reactors, so-called radiolytic splitting of water (H2O) molecules takes place in the reactor radiation field, that is to say this leads to the formation of so-called radiolysis gases. Particularly in the case of boiling water reactor (BWR) installations, continuous radiolysis gas formation can also occur in the core area, with the radiolysis gases that are produced in this case, hydrogen and oxygen, being transported to the turbine condenser in the steam that is in the main circuit of the BWR installation. In addition to such continuous releases of hydrogen, and such releases of hydrogen resulting from operation, the release of hydrogen gas and carbon monoxide must be expected within the safety vessel or containment, which surrounds the reactor core, in the event of fault or accident situations in a nuclear installation, in which oxidation of zirconium can occur, for example as a result of core heating. In this case, large amounts of hydrogen may be released, particularly after a fault involving coolant loss. Particularly during power operation of the installation, with radiolysis gases being formed continuously, the hydrogen that is produced may furthermore be enriched, for example in the area of the turbine condenser. As a result of which explosive gas mixtures occur within the containment or within the components in these operating or fault situations, with the enrichment of hydrogen in the respective atmosphere being possible to the extent that the integrity of the respective components could be endangered, particularly in the event of corresponding oxygen enrichment, in the case of accidental ignition by combustion of a relatively large amount of hydrogen since, at the start of the fault, the BWR containment is still reliably inerted over the first days. Various devices or methods are discussed in order to prevent the formation of explosive gas mixtures in the containment, or in other components of a nuclear installation, as a result of processes such as these. By way of example, these include devices such as catalytic recombiners, catalytically and/or electrically operating ignition devices or the combination of the two above-mentioned devices as well as methods for permanent or subsequent inerting of the containment. In the case of BWR installations, use of systems such as these may in this case be envisaged in conjunction with continuous purging in the turbine condenser area, since the concentration of gases that require treatment occurs predominantly in the condenser areas. When using a catalyst recombiner to remove the hydrogen from the atmosphere in the containment by its controlled oxidation with the assistance of a catalytic converter, the aim in particular is to achieve early and flameless recombination of the hydrogen with oxygen. In this case, a significant pressure buildup resulting from virulent hydrogen combustion should be reliably prevented. In order to allow this requirement to also be complied with safely with respect to the heat that is normally released as a result of the recombination or oxidation reaction, catalytic recombiners are normally configured exclusively for the conversion of gases with a comparatively low proportion of combustible gases considerably below the ignition limit, that is to say for example with a maximum proportion of the hydrogen carried therein of 4% by volume. When using catalytic recombination systems such as these for gas processing in the area of the turbine condenser of a nuclear power station, purging is normally provided in the turbine condenser area. The gas flow, which has hydrogen in it and requires treatment, is in this case supplied from the turbine condenser via an inlet line to a suitable recombination device in which a recombination reaction of the hydrogen carried in the gas with oxygen to form water is initiated in a reaction zone with a number of suitable selected catalytic converter elements. In order to ensure comparatively high operational safety in this case, initial inerting of the gas flow which requires treatment is normally provided in systems such as these before the gas flow enters the recombiner device, with an inerting level, which is suitably high to prevent an explosive gas reaction, being achieved by suitable addition of steam to the gas flow. In gas processing systems such as these, provision is made for steam to be added to the gas flow, before it enters the reaction zone, for this purpose. With regard to process-dependent fluctuations in the production rate of the radiolysis gases and other varying operating parameters, purging and gas treatment systems such as these are normally configured to ensure sufficiently high operational safety with sufficiently stringent safety margins, in particular relating to the intended proportion of steam in the gas flow after the steam addition. The feed rates of the dilution steam are in this case normally considerably greater than the actually expected requirement, particularly with regard to the stated stringent safety margins, with a hydrogen concentration in the gas flow being considerably less that that assumed for design purposes, in particular with regard to the fluctuating operating conditions, possibly in a large number of operating states. Furthermore, in the case of systems such as these, it must normally be accepted that the steam consumption will be considerably greater than the actual requirement. Because of the comparatively low hydrogen concentrations that occur, this large amount of steam consumption causes the recombination devices which are provided for gas processing to operate outside the actually expedient parameter ranges for reliable recombination operation. Furthermore, in order to prevent or reduce corrosion effects and to ensure a comparatively homogeneous hydrogen concentration in the flow medium of the main circuit, gassing with hydrogen can be provided in a nuclear installation in order to achieve a “Hydrogened Water Chemistry” (HWC) process which, in the end, leads to an additional reduction in the amounts of radiolysis gases carried in the coolant, as a result of the achievable high degree of homogenization of the gas components. Measures such as these make it possible to reduce the rate of formation of hydrogen in the system and thus the hydrogen enrichment in the turbine condenser to values of considerably less than 10% of the originally assumed design proportions, with the result that, in systems such as these, the recombination device which is connected in the purging and gas treatment system operates only comparatively unreliably because of the low reaction temperatures which now occur, as a consequence of which the reaction kinetics are slowed down to a major extent and there are low process overheating margins. It is accordingly an object of the invention to provide a method for catalytic recombination of hydrogen, which is carried in a gas flow, with oxygen, and a recombination system for carrying out the method which overcomes the above-mentioned disadvantages of the prior art devices of this general type. The invention is therefore based on the object of specifying a method of the abovementioned type for catalytic recombination of hydrogen, which is carried in a gas flow, with oxygen, by which, even in the case of the operating conditions or ways of operation, which vary in this case, a particularly high operational reliability of the recombination device is ensured, particularly with regard to hydrogen feed which is provided as required in the primary circuit of the installation. With the foregoing and other objects in view there is provided, in accordance with the invention, a method for catalytic recombination of hydrogen, being carried in a gas flow, with oxygen. The method includes the steps of adding steam to the gas flow before the gas flow enters a reaction zone, passing the gas flow through the reaction zone having a number of catalytic converter elements, and adjusting a feed rate of the steam to be added in dependence on a measured value being characteristic of a current actual temperature in the reaction zone. With regard to the method, the object is achieved according to the invention in that steam is added to the gas flow before it enters the reaction zone, and with the feed rate of the steam to be added being adjusted in dependence on a measured value being characteristic of a current actual temperature in the reaction zone. The invention is in this case based on the idea that, in order to ensure high operational reliability, the case of the maximum hydrogen release which can be expected is used as the basis for the configuration of a gas processing system of the stated type such that a sufficiently large amount of dilution steam is added in the systems that are normally provided, based on the maximum expected hydrogen release rate, such that, even in this extreme case, sufficiently high inerting of the gas flow supplied to the recombination unit is ensured. However, particularly with regard to the varying operating conditions and methods, this case corresponds to an extreme exceptional situation, as a result of which the added steam or dilution steam is fed to a considerably excessive extent to the system throughout the vast majority of the operating range of an installation such as this. Such overfeeding of this system with added steam, which goes well beyond the actual requirement, should, however, be avoided since, on the one hand, undesirably high steam consumption must be accepted as a result of such overfeeding and since, on the other hand, with such overfeeding, the proportion of hydrogen in the gas flow entering into the reaction zone in most operating situations of the installation is so low that the recombination unit operates outside its stable operating parameters, and therefore only unreliably. In order to counteract this, the addition of steam to the gas flow should be appropriate for the requirement and should be oriented on the actual current operating conditions. When steam is added as required in this way, a particularly reliable functionality in the respective operating state of the recombination device which is provided for the recombination reaction should be provided as a design aim. Particularly with regard to the specific reaction conditions for the conversion of hydrogen and oxygen to water and with regard to the catalytic converter materials which are normally used, such as preferably platinum and/or palladium, the aim in this case should be to set an operating temperature range which is considered to be particularly advantageous in the reaction zone, preferably from about 300° C. to 600° C. In order to achieve this design aim, that is to say to set a temperature level which is particularly advantageous for the recombination reaction in the reaction zone, even when the hydrogen concentrations in the gas flow are varying, the steam feed or steam addition should be controlled or restricted via the temperature in the reaction zone. In this case, the particularly desirable temperature level from about 300° C. to 600° C. in the reaction zone can be set, in particular, by preferably setting a hydrogen content in the gas flow flowing into the reaction zone to be about 3 to 8% by volume. The steam feed is therefore advantageously controlled or set such that a hydrogen concentration such as this occurs in the gas flow. However, in order to also guarantee sufficiently high operational safety in any case here as well, an additional or alternative advantageous refinement provides for the proportion of steam in the gas flow entering the reaction zone to be set to at least 70% by volume. This ensures adequate inerting of the gas flow flowing into the reaction zone in all circumstances. In this case, the steam that is added can essentially be set on the basis of a suitable temperature characteristic value in the reaction zone as a reference variable. However, in order to allow the steam feed to be matched to the changing operating states to match the particular requirement, the setting of the steam flow to be added is in one advantageous refinement set taking account of a plurality of temperature characteristic values in the reaction zone and/or taking account of the time profile of one or more of the parameters. In this case, in particular, it is possible to take account of the temperature profile within the reaction zone or the temperature difference at its inlet and at its outlet. Furthermore, if the development of the parameters over time is taken into account in the form of a predictive setting, the system parameters can be readjusted in a particularly timely form. In this case, in particular, it is possible to take account of the inlet temperature and the outlet temperature in the reaction zone, a temperature difference determined from this, the pressure loss of the gas flow in the reaction zone, and/or the time profile of one or all of these parameters. The addition of steam into the gas flow in dependence on the temperature in the reaction zone makes it possible to take particular account, in particular based on the requirement, of the varying operating conditions and states in the turbine condenser of a nuclear power station. The gas flow is therefore advantageously taken from the turbine condenser of a nuclear power station. With regard to the recombination system, in which an inlet line, which is provided in order to supply the gas flow into a reaction zone which is formed by a number of catalytic converter elements is connected to a steam supply line, in order to add steam as required. The stated object is achieved in that a control unit, which is associated with a metering valve connected in the steam supply line, is connected on the data input side to a number of temperature sensors associated with the reaction zone. The control unit is in this case advantageously in the form of a regulator unit with which a temperature value of the reaction zone is associated as a reference variable. In a further advantageous refinement, the control unit is provided with a data memory in which the determined measured values, that is to say in particular the temperature characteristic values and/or the pressure or partial pressure values, are stored and are kept available for subsequent evaluation of the time profile of the stated characteristic values. The inlet line is in this case advantageously connected on the input side to the turbine condenser of a nuclear power station. Particularly in an arrangement such as this, the recombination system is actually also additionally suitable for use as an auxiliary installation, as required, for handling fault scenarios in the reactor safety vessel or containment. In order to allow this to be done in a particularly simple manner, in one particularly advantageous refinement, the inlet line is connected via a branch line to the safety vessel of the nuclear power station. Therefore, in the event of faults with hydrogen being released in the containment area, the recombination device can be used for oxidation or recombination of the amounts of hydrogen to be expected in the containment, or in principle also, by feeding in hydrogen, to produce or maintain containment inerting using nitrogen. In the case of use as required such at this, the recombination device is advantageously operated primarily in a sliding form on the circuit, on the respectively prevailing pressure level in the containment. The advantages achieved by the invention are in particular that, the setting of the feed rate of the steam to be added as a function of the temperature in the reaction zone allows steam to be fed in in a manner which is particularly matched to the requirements, thus conserving resources. Specific steam feeding such as this thus also makes it possible to take account of the varying hydrogen proportions in the gas flow that requires treatment, in such a way that, even in normal operating conditions and in the operating mode in which hydrogen is injected (HWC mode) into the steam feed water circuit, a hydrogen concentration of preferably 3 to 6% by volume can be set at the inlet to the recombination zone on the basis of the amount of steam that is added being then correspondingly reduced. Even in the stated operating conditions, this therefore makes it possible to apply to the recombination device a hydrogen concentration which is particularly advantageous for the recombination reaction, and thus to maintain operating conditions which are particularly advantageous for the recombination reaction. In this case, in comparison to conventional installations, it is possible to achieve a saving in the amount of dilution steam or the amount of propellant steam by a factor of up to 10, and a significant reduction in the operating costs associated with this. Furthermore, with regard to the reduced amounts of steam, it is possible to achieve a reduction in the pipelines and other components that are used, with the life of the components that are used furthermore still being increased by the comparatively lower mechanical loads. This method and the device therefore allow and introduce a highly effective “Efficiency Controlled Recombination Process Technology”, referred to for short as ECO REC technology. 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 method for catalytic recombination of hydrogen, which is carried in a gas flow, with oxygen, and a recombination system for carrying out the method, 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. The construction and method of operation of the invention, however, together with additional objects and advantages thereof will be best understood from the following description of specific embodiments when read in connection with the accompanying drawings. Identical parts are provided with the same reference symbols in the two figures. Referring now to the figures of the drawing in detail and first, particularly, to FIG. 1 thereof, there is shown a recombination system 1 that is configured for catalytic recombination of a combustible gas, that is to say hydrogen in the exemplary embodiment. In the exemplary embodiment, the recombination system 1 is in this case intended for gas processing or treatment for a turbine condenser 2 of a nuclear power station. For this purpose, an inlet line 4 of the recombination system 1 is connected on an input side, in the form of a purging system, to the turbine condenser 2 of the nuclear power station. On an output side, the inlet line 4 into which a primary compressor 6 and a secondary compressor 8 are connected in order to feed the gas flow G which requires treatment, is connected to a recombination unit 10. The recombination unit 10 is in this case configured for the actual catalytic recombination reaction in order to remove hydrogen which is carried in the gas flow G. For this purpose, a reaction zone 12 is provided within the recombination unit 10, and is formed by a number of catalytic converter elements. In this case, the catalytic converter elements may be configured in different ways in embodiments which are known per se, in which case, in particular, it is possible to use a configuration as plate or film catalytic converters, or else as a catalytic converter filling. In the exemplary embodiment, a first catalytic converter 14 is provided in the reaction zone 12, formed from suitable selected plate elements composed of palladium and/or platinum, preferably as a washcoat on metallic supporting bodies, which is arranged via a flow distribution and alignment system 16 on a catalytic converter filling 18, once again based on platinum and/or palladium. In order to promote suitable reaction conditions, an outer casing 20, which surrounds the reaction zone 12, of the recombination unit 10 is provided with casing heating 22. On the output side, the recombination unit 10 is connected to an outlet line 24, via which the treated and processed gas flow can be carried away. The outlet line 24 in this case opens in an off-gas chimney 30 via a drying device 26 and a deceleration path 28, preferably based on activated charcoal. In order to ensure a particularly high level of operational safety in the recombination of the hydrogen fraction in the gas flow G, the recombination system 1 is configured such that the recombination reaction can be carried out in inert conditions, and thus precluding any risk of explosion. To this end, the addition of steam is envisaged as a measure for making the gas flow G inert before it enters the recombination unit 10. In order to add steam D as required, the inlet line 4 is in this case connected to a steam supply line 32. In order to recover the steam D that is fed in in this case on the output side of the recombination unit 10 and in order to make it available as water once again to the installation process, a number of condensers 34, 36 are connected in the outlet line 24, in which the steam component which is carried in the treated gas flow is condensed to form water. On the water side, the condensers 34, 36 are connected via a return flow line 38 to the turbine condenser 2. The water which is fed back in this way via the return flow line 38 into the turbine condenser 2 is passed from this into the primary circuit of the nuclear power station, which is indicated in FIG. 1 by a feed water line 40 and a feed water pump 42. The recombination system 1 is configured for particularly high operational reliability of the recombination device 10 even in varying operating conditions and with varying hydrogen fractions in the gas flow G which requires treatment. This takes particular account of the fact that a fluctuating release of hydrogen as a radiolysis gas can be expected in the turbine condenser 2 as a result of varying operating conditions during regular operation. Furthermore, this takes account of the fact that the nuclear power station in the exemplary embodiment is configured, based on requirement or regularly, with a hydrogen feed in the primary circuit, as is indicated by the feed line 44, by which the corrosion resistance of the overall system is increased and the overall hydrogen rates that can be observed can be reduced as a result of the greater homogenization that can be achieved in the primary coolant. In order to correspondingly also ensure adequately high catalytic activity in the recombination device 10 at the comparatively low hydrogen rates that can normally be expected, and thus to ensure particularly high operational reliability, the recombination system 1 is configured for steam to be added to the gas flow G as required and depending on the situation. In this case, a first intermediate station 46 is connected in the steam supply line 32, via which a pressure of preferably more than five to ten times the critical pressure at the Laval velocity is set in the downstream section of the steam supply line 32. A further pressure reducing device, which can be regulated, or a metering valve 48, which can be regulated, is then connected in the steam supply line 32. The metering valve 48, which can be regulated or can be controlled, is in this case driven via an associated control unit 50. This configuration results in the gases being purged from the turbine condenser 2 by steam jet pumps in the form of compressors 6, 8 with the gas flow G simultaneously being diluted with steam, with the final compression stage, seen in the flow direction, that is to say the compressor 8, being in the form of a regulated steam jet stage. The propellant steam in the propellant nozzle inlet is kept above the critical pressure ratio over a wide operating range by suitably driving the metering valve 48 in this steam jet stage. The jet inlet pressure range of this stage is in this case chosen to be at least sufficiently high that the critical pressure ratio is exceeded by a factor of 3 to a factor of 10 with respect to a downsteam pressure of about 1 bar. In particular, this ensures that, even in the case of a major intermediate restriction in the jet pump propellant nozzle, the advantageous Laval velocity is always ensured, with a correspondingly advantageous feed capability for the gas flow. By way of example, with an intermediate restriction of 15 bar to about 2 bar, the amount of steam can be reduced by factor of more than 7 while nevertheless still achieving the high velocities, up to the Laval velocity, which are important for feed purposes, with an expansion to 1 bar in the nozzle. For the addition of steam in particular as required even in varying operating conditions, the recombination system 1 is configured with the design aim of operating the catalytic converter elements in the reaction zone 12 largely in a temperature range of, for example, between 300° C. and 600° C., which is particularly advantageous for the recombination reaction. In order to make this possible, provision is specifically made for a suitable amount of steam to be added into the gas flow G such that the desired temperatures are achieved in the reaction zone as a consequence of the resultant mixing and concentration ratios. In order to make this possible, the control unit 50 is in the form of a regulator unit with a temperature value of the reaction zone 12 as a reference variable and, on the data input side, is connected to a number of temperature sensors 52, 54, 56 which are associated with the reaction zone 12. In this case, a first temperature sensor 52 is arranged in the gas inlet area of the reaction zone 12, and a second temperature sensor 54 is arranged in the gas outlet area of the reaction zone 12. A third temperature sensor 56 is, in contrast, arranged in the interior of the reaction zone 12. The measured values provided by the temperature sensors 52, 54, 56 therefore make it possible to take account of suitable temperature characteristic values in order to ensure the desired temperature level for the steam feed. Furthermore, other parameters such as temperature profiles or else the time response of the individual temperature characteristic values, or else the difference temperature between the gas inlet and the gas outlet, can also be evaluated and taken into account in a suitable form. Furthermore, the control unit 50 is provided with a data memory 91, such that the measurement values can be suitably temporarily stored and can be evaluated further as required. For further system monitoring, further sensors, in particular material sensors, are connected to the individual lines, in particular to the outlet line 24. In this case, by way of example, an oxygen sensor 58 can be used to detect a lack of oxygen in the off-gas flow, which can be compensated for in a suitable form, as required, via a feed run 60 for oxygen. A hydrogen sensor 95 is additionally connected between the outlet line 24 and the control unit 50. Furthermore, the control unit 50 is configured to ensure adequate inerting of the gas flow G, before it enters the recombination unit 10, during every situation during operation. To this end, the control unit 50 is configured to set the feed rate of the steam D to be added such that the gas flow G has a steam content of at least 70% by volume when it enters the reaction zone 12. The recombination system 1 is therefore configured for particularly flexible addition of steam as required into the gas flow G, thus ensuring reliable gas treatment of the gas purged from the turbine condenser 2. Furthermore, however, the recombination system 1 is also configured for connection as required during the treatment of fault scenarios in which hydrogen is released in the safety vessel or containment 70 of the nuclear installation. To this end, the inlet line 4 of the recombination system 1 is connected to the containment 70 via a branch line 72, which opens into the inlet line 4 at a feed point 74. In a fault scenario such as this, in which a medium which requires treatment is fed into the recombination device 10 via the branch line 72, an operating mode is envisaged essentially in the equal-pressure mode at a pressure in the circuit which corresponds to the respectively prevailing containment pressure level. To this end, a return line 76 is connected to the outlet line 24, via which return line 76 circulating operation can be produced, in exchange with containment 70, when required. FIG. 2 shows one alternative embodiment of the recombiner device 10. In this case, in one particularly advantageous refinement, the reaction zone 12 is arranged within an inner casing, which is integrated in an outer casing 82. The inlet area between the reaction zone 12 and the outer casing is heated by a heating device which is fitted directly to the outer casing 82, such that the desired temperatures can be maintained. Furthermore, an additional heater 84 is provided in the inlet area for the gas flow G. In the exemplary embodiment shown in FIG. 2, the reaction zone 12 is followed by a spray system 90 with a number of spray nozzles 92, to which cooling water can be supplied via a supply line 94. The spray system 90 can thus be used for injection cooling or spray cooling of the gas flow flowing out of the reaction zone 12, thus allowing recooling of the treated gas flow even without downstream condensers. An arrangement such as this thus allows a particularly compact construction within a frame structure 100 such that it is possible to retrofit existing installations, in particular even when very little space is available. The separation of the pressure vessel casing 82 from the reaction zone 12 by an annular gap and the direct subsequent spray cooling furthermore result in the high temperatures in the reaction zone 12 being decoupled from the pressurized outer area, thus also allowing reaction temperatures of more than 500° C.—without any direct effects on the materials, including the limiting characteristic data of the pressurized part. In addition, recuperative gas heating by the high-temperature section and casing outer heating is provided in the system shown in FIG. 2. |
|
060312378 | abstract | A radiation image storage panel has a phosphor layer comprising a stimulable phosphor and a binder, in which the binder is composed of a resin containing a thermo-plastic polyurethane elastomer and a radical scavenger. The panel shows excellent durability against both light and repeated conveying. |
summary | ||
summary | ||
description | This application claims priority to a provisional patent application by the same inventors, entitled: “Statistical Performance Monitoring,” Ser. No. 60/419,175, filed on Oct. 17, 2002. This application is related to an application by the same inventors, entitled: “Enterprise Management System and Method which Includes Statistical Recreation of System Resource Usage for More Accurate Monitoring, Predication and Performance Workload Characterization,” Ser. No. 09/287,601, filed on Apr. 7, 1999. Both of the above applications are incorporated herein by reference. 1. Field of the Invention This invention relates generally to system performance monitoring, especially for performance monitoring of a distributed computer network system with a massive number of nodes or consoles. 2. Description of the Related Art The data processing resources of business organizations are increasingly taking the form of a distributed computing environment in which data and processing are disbursed over a network comprising many interconnected, heterogeneous, geographically remote computers. Such a computing environment is commonly referred to as an enterprise computing environment, or simply an enterprise. Managers of the enterprise often employ software packages known as enterprise management systems to monitor, analyze, and manage the resources of the enterprise. Enterprise management systems may provide for the collection of measurements, or metrics, concerning the resources of individual systems. For example, an enterprise management system might include a software agent on the individual computer system for the monitoring of particular resources such as CPU usage or disk access. U.S. Pat. No. 5,655,081 discloses one example of an enterprise management system. In a sophisticated enterprise management system, tools for analysis, modeling, planning, and prediction of system resources utilization are useful for assuring the satisfactory performance of one or more computer systems in the enterprise. Examples of such analysis and modeling tools are the “ANALYZE” and “PREDICT” components of “PATROL Perform/Predict for UNIX or Windows” or “BEST/1 for Distributed Systems” available from BMC Software, Inc. Such tools usually require the input of periodic measurements of the usage of resources such as CPUs, memories, hard disks, network bandwidth, number of files transferred, number of visitors to a particular web page, and the like. To insure accurate analysis and modeling, therefore, the collection of accurate performance data is critical. Many modern operating systems, including “Windows NT” and UNIX, are capable of producing an enormous amount of performance data and other data concerning the state of the hardware and software of the computer system. Such data collection is a key step for any system performance analysis and prediction. The operating system or system software collects raw performance data, usually at a high frequency, stores the data in a registry of metrics, and then periodically updates the data. In most case, metric data is not used directly, but instead sampled from the registry. Sampling at a high frequency can consume substantial system resources such as CPU cycles, storage space, and I/O bandwidth. Therefore, it is impractical to sample the data at a high frequency. On the other hand, infrequent sampling cannot capture the complete system state: for example, significant short-lived events and/or processes can be missed altogether. Infrequent sampling may therefore distort a model of a systems performance. The degree to which the sampled data reliably reflects the raw data determines the usefulness of the performance model for system capacity planning. The degree of reliability also determines the usefulness of the performance statistics presented to system managers by performance tools. Sensitivity to sampling frequency varies among data types. Performance data can be classified into three categories: cumulative, transient, and constant. Cumulative data is data that accumulates over time. For example, a system CPU time counter may collect the total number of seconds that a processor has spent in system state since system boot. With transient data, old data is replaced by new data. For example the amount of free memory is a transient metric which is updated periodically to reflect the amount of memory not in use. For transient metrics the only way to find even approximate means, variances, or standard deviations is to do periodic sampling. The third type of performance data, constant data, does not change over the measurement interval or lifetime of the event. For example, system configuration information, process ID, CPU model type, and process start time are generally constant values. Of the three data types, transient performance metrics are the most sensitive to variations in the sampling interval and are therefore, the most likely to be characterized by uncertainty. For example, with infrequent sampling, some state changes may be missed completely. However, cumulative data may also be rendered uncertain by infrequent sampling, especially with regards to the calculation of the variation of such a metrics. Clearly then, uncertainty of data caused by infrequent sampling can cause serious problems in performance modeling. A related patent application titled “Enterprise Management System and Method Which Include Statistical Recreation of System Resource Usage for More Accurate Monitoring, Prediction and Performance Workload Characterization,” Ser. No. 09/287,601, discloses a system and method that meets the needs for more accurate and efficient monitoring and prediction of computer system performance. Even when sampling frequencies are reduced, the performance data collected by system monitors can still be enormous. Traditional performance monitoring methods and/or tools display performance metric values at a rate similar to the rate they are sampled. To accurately monitor the hardware and software of a computer system, many different metrics are sampled, collected, stored and/or reported. When a computer network system or enterprise comprises only a few nodes, the aggregation of the monitoring data from each of the few nodes may not be a problem. But when the system grows, the performance data collected from each computer or node will increase proportionally. The large quantity of data that has to be pushed or pulled across a network for displaying or reporting becomes impractical or even impossible when hundreds or even thousands of nodes are managed from a few nodes or consoles. Therefore, it is desirable to have a method or system to further reduce the growth of data quantity in order to maintain the ability to monitor the performance of each node. The present invention uses statistical parameters, such as mean, standard deviation, and exceptional value to reduce the amount of system performance data collected and transmitted to a system performance monitor for system performance monitoring and analysis. In one embodiment, to reduce the amount of data collected for analysis, appropriate metrics are selected for different system performance monitoring; appropriate thresholds or ranges for the metrics are set; the data collection frequencies may also be varied depending on the metrics used. Sampled data for a particular performance metric within a range are not reported, but are replaced with the average of the metric. Only the data that are outside the range or threshold are reported for analysis and/or visualization. In another embodiment, the average of the metric is updated constantly by the Collector. When at the end of a measurement period the updated average differs from the original average (that was being used by the system performance monitor) by an amount that exceeds a threshold, then the new average replaces the old average. The new average is stored and reported to the system performance monitor. In a third embodiment, various metrics are compared and their inter-dependences are determined. If the correlation between two metrics is within a certain range or threshold, then only the first metric is collected, transmitted and reported for both metrics. Thus the number of metrics needed to be monitored is decreased without losing any important information. FIG. 1 illustrates an enterprise computing environment. The enterprise 100 comprises a plurality of computer systems which are interconnected through one or more networks. One or more local area network (LANs) 104 may be included in the enterprise 100. A LAN 104 is a network that spans a relatively small area. Typically, a LAN 104 is confined to a single building or group of buildings. Each node (i.e., an individual computer system or device) on a LAN 104 preferably has its own CPU with which it executes programs, and each node is also able to access data and devices anywhere on the LAN 104. The LAN 104 thus allows many users to share devices as well as data stored on file servers. The LAN 104 may be characterized by any of a variety types of topology (i.e., the geometric arrangement of devices on the network), of protocols (i.e., the rules and coding specifications for sending data, and whether the network uses a peer to-peer or client/server architecture), and of media (e.g., twisted pair wire, coaxial cables, fiber optic cables, radio waves). As illustrated in FIG. 1, the enterprise 100 includes one LAN 104. However, the enterprise 100 may include a plurality of LANs 104 which are coupled to one another through a wide area network (WAN) 102. A WAN is a network that spans large geographic areas. Each LAN 104 comprises a plurality of interconnected computer systems and optionally one or more other devices: for example, one or more work stations 110a, one or more personal computers 112a, one or more laptop or notebook computer systems 114, one or more server computer systems 116, and one or more network printers 118. As illustrated in FIG. 1, the LAN 104 comprises one of each computer systems 110a, 112a, 114, and 116, and one printer 118. The LAN 104 may be coupled to other computer systems and/or devices and/or LANs 104 through a WAN 102. One or more mainframe computer systems 120 may optionally be coupled to the enterprise 100. As shown in FIG. 1, the mainframe 120 is coupled to the enterprise 100 through the WANT 102, but alternatively one or more mainframe 120 may be coupled to the enterprise 100 through one or more LANs 104. As shown, the mainframe 120 is coupled to a storage device or file server 124 and the mainframe terminals 122a, 122b, and 122c. The mainframe terminals 122a, 122b, and 122c access data stored in the storage device or file server 124 coupled to or comprised in the mainframe computer system 120. The storage device can also couple to LAN, WAN, Internet and/or computer systems of different platforms. The enterprise 100 may also comprise one or more computer systems which are connected to the enterprise 100 through the WAN 102: as illustrated, a workstation 110b and a personal computer 112b. In other words, the enterprise 100 may include one or more computer systems which are not coupled to the enterprise 100 through LAN 104. For example, the enterprise 100 may include computer system which are geographically remote and connected to the enterprise 100 through the internet. To manage or monitor the performance of the network enterprise network system 100, some of the computers in the network for example, 110d as shown in FIG. 2 may act as a monitor or management console. The management monitor 110d will request and receive various performance measurement data from all the computers within the network system. With the various different performance data or metrics collected from the various computers connected to the network system 100, the monitor 110d can perform analysis on the performance of those various computer connected to the enterprise 100. When the enterprise system 100 has only a few nodes or even a few dozen nodes, the data collection for the performance analysis will not burden the network excessively. But when the number of nodes increases into hundreds or even thousands, the amount of data related to the system performance measurement collected at each node, forwarded to the monitor 110d may become prohibitively large. One of the benefits of the current invention is to reduce substantially the amount of data transferred from each node to the monitoring node. FIG. 3 shows an overview of the enterprise management system 180. The enterprise management system 180 includes at least one console node 400 (such as monitor 110d discussed above) and at least one agent node 300, but it may include a plurality of console nodes 400 and/or a plurality of agent nodes 300. In general, an agent node 300 executes software to sample/collect metric data on its computer system 150, and a console node 400 executes software to monitor, analyze, and manage the collected metrics from one or more agent nodes 300. A metric is as measurement of a particular system resource. For example, the enterprise management system 180 collects metrics such as CPUs, disk I/O, file system usage, database usage, thread, processes, kernel, registry, logic volumes, paging, number of visitors to a web page, pages viewed, types of web browsers. Each computer system 180 in the enterprise 100 may comprise a console node 400, an agent node 300, or both a console node 400 and an agent node 300. The console node 400 may comprise four user visible components: a monitor component 402, a collect graphical user interface (GUI) 404, and Analyze component 406, and a Predict component 408. Both Analyze and Predict components have their GUI as well. All four components 402, 404, 406, and 408 of the console node 400 may be part of the “Perform/Predict for UNIX or Windows” or “BEST/1 for Distributed Systems.” software package or for the “PATROL” software package, or available from BMC Software, Inc. The agent node 300 may comprise an agent 302, one or more data collectors 304, universal data repository (URD) history files 210a, and universal data format (UDF) history files 212a. The agent node 300 may include either of UDR 210a or UDF 212a, but not both. The monitor component 402 allows a user to monitor, in real time, data that is being collected by an agent 302 and being sent to the monitor 402. The collect GUI 404 is employed to schedule data collection on an agent node 302. The analyze component 406 takes historical data from a UDR to 102A and/or UDF 212 to create a model of the enterprise 100. The predict component 408 takes the model from the analyze component 406 and allows a user to alter the model by specifying hypothetical changes to the enterprise 100. Analyze 406 and Predict 408 can create output in a format which can be understood and displayed by a Visualizer 204. Agent 302 controls data collection in a particular computer system and reports the data in real time to one or more monitors 402. The data collectors 304 collect data from various processes and subsystems of the agent node 300. The agent 302 sends real time data to UDR 210A, which is a database of historical data in a particular data format. The UDF 212a is similar to that UDR 210a, but the UDF 212a uses an alternative data format and is written directly by the data collector 304. FIG. 4 shows an overview of the monitor component 402 of the console node 400 of the enterprise management system 180. The monitor 402 comprises a manager daemon 430, one or more monitor consoles (as illustrated, 420a and 402b), and a policy registration queue 440. Although two monitor consoles 420a and 420b are shown in FIG. 4, there may be one or more consoles executing on any of one or more console nodes 400. FIG. 5 shows a typical agent component 302 of the agent node 300 of the enterprise management system 180. Every agent node 300 has one agent 302. The monitor console 420c is another instance of the monitor consoles illustrated in FIG. 5 with reference number 420a and 420b. When a user desires to start an agent 302 and begin collecting data on a particular agent node 300, the user operates the monitor console 420c to issue an agent star request through a service daemon 202b. The service daemon 202b is always executing on the agent node 300 in order to intercept messages from one or more monitor consoles 420 even when the agent 302 is offline. The service daemon 202b also intercepts agent version queries from the monitor console 420c. The monitor console 420c may also send a collection request, which requests the agents 302 to begin collecting particular metrics or metrics groups on the agent node 300. When the agent 302 receives a collect request from the monitor console 420c through the service daemon 202b, the agent 302 initiates the collection through the collect registry queue (CRQ) 340. The agent 302 uses the CRQ 340 to control and schedule data collection. By helping the agent 302 know how many collectors 304 are running and whether the collector 304 are each the right type, the collect registry queue 340 prevents redundant collection. After metrics data is collected, the data is transferred to a metrics repository 350. The metrics repository 350 sits between the agent 302 and the collectors 304 and provides fast communication between the agent process 302 and the collector processes 304. According to one embodiment of the current invention, rather than reporting all the collected metrics data from the agent 302 to the monitor console 420 as in some prior art methods, the metrics data are processed by the agent 302 and to reduce the amount of data that needs to be reported. One method according to the current invention to reduce the amount of data collected and stored and transferred between agent 302 and monitor console 420 is to use statistical performance monitoring. The focus of this method is on combining statistics of metrics for a larger interval, rather than retaining metrics at sample interval level. Performance metric values are often sampled every few seconds. This generates huge amounts of data when a system is monitored continuously with many metrics. For instance, at a five second sampling interval, 17,280 data points will be collected in just twenty-four hours and that is for only one metric. Systems may have over 100 metrics which means that the thousands of nodes will generate billions of data points each day. This is too much, especially since most of the data may not be interesting. According to the methods of some embodiments of the current invention, the uninteresting data or data with redundant information are filtered out. The data is not needed if it is within a “boring” range. A value can be defined to be “boring” in many different ways. For instance, 1) if the difference of the sampled value and the average is within the standard deviation. In this case, both first moment (the average) and second moment (the standard deviation) are calculated; 2) if the difference is within some percentage, e.g. 20% of the average. In this case, only the first moment (the average) is calculated; or 3) if the difference is within a user defined range of the average, for example any value less than 100. In this case, the range or threshold is not related to the present sampled data, but based on historical or empirical data. With this method, for metrics of interest, when the sample is within the boring range, the data is not reported and the system performance monitor assumes the data is the average. When the sample is outside the boring range, or “interesting”, then it is collected and reported. From a statistical point of view, as an example, if a metric is sampled at a 5-second interval, and summarized and spilled every 15 minutes, the average obtained for the 15-minute spill has a possible error of about 19% at a 99% confidence interval. That is, we can be 99% certain that the error is no more than 19%. The following is a brief explanation of the relationship between the errors, confidence level, the number of samples collected and their averages. According to the central limit theorem the c% confidence interval for the metric population is from{overscore (x)}−f(c)s/√{square root over (n)} to {overscore (x)}+f(c)s/√{square root over (n)} (1) where {overscore (x)} is the sample mean, s is the sample standard deviation, n is the number data in the sample, and f(c) is the (1+c/100)/2-quantile of the unit normal distribution. One can find f(c) in most statistics books. A few examples are listed in table 1. TABLE 1Four confidence intervals, 80%, 90%, 95%, 99%,and their 0.90- 0.95-, 0.975-, 0.995-quantile of the unit normal distribution.ConfidenceInterval c(1 + c/100)/2f(c)80%0.91.28290%0.951.64595%0.9751.96099%0.9952.576 Assume that the sample mean is off by +e% from the metric population mean. From (1) we have{overscore (x)}+f(c)s/√{square root over (n)}={overscore (x)}(1+e/100) (2) Let C=s/{overscore (x)} be the coefficient of variation of the sample. Then, from (2), error percent, e%, could be represented in terms of sample size n, C, and f (c): ⅇ = 100 n f ( c ) C n or n = ( 100 f ( c ) C ⅇ ) 2 ( 3 ) In the case of a 5-second sample interval, the error percent of the average for the 15-minute spill would be: 100 180 f ( 99 ) C 180 = 19.2 % The above formula [0047] implies that the confidence interval is 99% and the data values are exponentially distributed, i.e., C=1. In other words, we are 99% sure that the true average (population average) for the 15-minute spill is within +/−19.2% of the computed average. It is quite clear that, because of the uncertainty inherited from the sampling process, storing, transmitting and reporting the interesting values of performance metrics make statistical sense. Formula (3) could likewise be used to determine the boring range based on the sample size and sample coefficients of variation for a given confidence interval. Note also that for the same sample size, n, and confidence interval, c, the variance would be off by ev percent: e v = 100 f ( c ) 2 n ,which is normally much less than the error for the mean. For the example given above, the variance would be off by only 100 f ( 99 ) 2 180 = 3.7 % . In general, the relationship between e and ev is: e v = n f ( c ) n C e , where C is the coefficient of variation of the data. Most performance models and modeling formulas only use averages. For instance, the key performance inputs for the models, such as workload throughputs, service times and utilization at servers are average numbers. So are the outputs of the models/formulas. For some more sophisticated modeling formulas, the first two moments may be used. As it is well known to the person skilled in the relevant art, the first moment {overscore (x)} of a sample is simply the average of the sample. A second moment {overscore (x2)} is simply the average of the squared values of the sample. With the first moment and the second moment, the standard deviation may be calculated. Third moment or above are very rarely used. Therefore, in most cases, mean and variance will be enough. The average referred through out this application may be many different kinds of average, including at least arithmetic or geometric averages, past static averages or running averages including current data, straight averages or weighted averages where some data are more important than others. The averages used by the methods in the current invention may be any one of them, or some combination of them. Different type of averages may be appropriate for different types of metrics with different data distributions. For example, when a given metric has a very large range, then geometric averages may be more appropriate than arithmetic averages. However, for most metrics, arithmetic average may be most appropriate. One useful average is an hour-by-hour straight average as used in the above example. An alternative is to compute a moving mean over multiple hours, with greater weight assigned to recent hours. A third alternative is to use historical data as well. For instance, average the previous hour with the current hour yesterday. Perhaps the most accurate alternative is to determine how closely the current hour yesterday matched the previous hour yesterday and use that relationship to adjust the average of the previous hour today. The closer the average used is to the real/true mean, the fewer exceptional values have to be reported, which means there will be less data to transmit or store. To obtain a closer average, a running average may need to be maintained and updated regularly. When the current running average differs from the original average by an amount greater than a threshold, the new running average will be reported/transmitted from the agent to the monitoring console. Thus, using a smaller threshold will cause more updated averages to be transmitted. The number of data points (sample size) that are needed, given an error range or boring range (mean +/−e%), to make the sampled average within a certain confidence interval, c, to the population average can be determined by formula (3) above. Another average is the Exponential Moving aVerage (EMV), which is a moving average that may give a greater weight to the latest data. The impact of old data gradually decreases. The current or n'th EMV, denoted by {overscore (d)}n(w), is based on the previous or (n−1)'th EMV, {overscore (d)}n−1(w) and the new or n'th data dn:{overscore (d)}n(w)=dnw+{overscore (d)}n−1(1−w),where w is a predefined weight, which may be any real number between 0 and 1 inclusive. The most obvious weight to choose is w = w f N where N is the moving window size and wf is a weight factor, which is any real number. When wf is less than 1, then the current data weighs less than the older data. With wf=2, the weight of the current data point is twice as important as the previous data point, etc., although a smaller scaling, say wf=1.3 may be more appropriate for a given metric. If wf=1 and N=n, then w = 1 n ,the EMV becomes the straight running average, i.e., {overscore (d)}n(w)={overscore (d)}n. For real-time monitoring the average is likely to be updated over time (e.g., using the EMV) rather than computed with all the data points collected so far. The same is true for computing variance as well. The following are two algorithms for updating the average and variance: Incremental update of average (mean): a process of computing current average, {overscore (d)}n, with a previous average, {overscore (d)}n−1, and a new data point, dn. The current straight running average can be computed by d _ n = d _ n - 1 + d n - d _ n - 1 n = [ ( n - 1 ) d _ n - 1 + d n ] / n Incremental update of variance: a process of computing current variance, σn2, with a previous variance, σn−12 and a new data point, dn. The current variance can be computed by the Sn/n Sum of variance = S n = ∑ i = 1 n ( d i - d _ n ) 2 Sn=Sn−1+(n−1)(dn−{overscore (d)}n−1)2/nσn2=Sn/n. Once the average and standard deviation are determined, the boring range may be selected. The selection of the “boring range” and the size of it will determine the amount of reduction in monitoring data collected, stored and/or transferred. The larger the range of the “boring range,” the fewer of data become “interest” and get transmitted from agent to console, the greater in the reduction of data transmitted. Quantitatively speaking, the less varying the data is, the fewer numbers need to be recorded. One could use a reliability function, R(x) [which is defined to be P(X≧x)], if one knows the distribution. For most of the common (non-power-tailed) distributions, P(X≧x) decays exponentially. The power-tailed distribution can be detected using the methods presented in U.S. Pat. No. 6,564,174, entitled “Enterprise management system and method which indicates chaotic behavior in system resource usage for more accurate modeling and prediction.” It is incorporated herein by reference. That means that the amount of data that needs to be collected/transmitted decreases drastically as the thresholds go up, i.e., defining a wider boring range. For example, assuming that the value of a performance metric is exponentially distributed, i.e., its distribution function, F(x), is:F(x)=1−e−λx, 0≦x<∞. Therefore, P(X≧x)=1−F(x)=e−λx. So, if one let x to be (mean+standard deviation), then only about 14 percent of data points needs to be stored. If x is (mean+2 times the standard deviation), then only about 5 percent of data points needs to be kept. See Table 3 below. Even if one does not assume any underlying distribution for the performance metrics, one can use Chebyshev's inequality to estimate the reduction in data volume. P ( X ≥ x ) ≤ σ 2 σ 2 + x 2 , ( 4 ) where σ2 is the variance. Formula (4) is distribution independent. One drawback is that it does not have a very tight upper bound. Table 2 shows some examples with a normal distribution. Table 3 shows an example for exponential distribution in which the tail of the distribution reduces much more slowly and for which the Chebyshev's upper bound is a little tighter. TABLE 2The Probability of a particular sample valueexceeds a predefined threshold x for anormal distribution.Chebyshev'sxP(X ≧ x)Upper BoundMean + σ15.9%50%Mean + 2σ 2.3%20%Mean + 3σ0.13%10% TABLE 3The Probability that a particular sample valueexceeds a predefined threshold x for anexponential distribution.Chebyshev'sxP(X ≧ x)Upper BoundMean + σe−2 = 13.5%20%Mean + 2σe−3 = 5.0% 10%Mean + 3σe−4 = 1.8% 5.9% Usually, only large values are “interesting.” Since, in general, half the values that differ from the mean by a large amount are small values, significant additional savings can occur by only storing large values that exceed the threshold. When only large values are of concern, the boring range can be defined as 0 through (Mean+3σ). In operation according to an embodiment of the current invention, when a system metric is to be monitored and analyzed for system performance for a node 300, an agent 302 will collect samples of the metric for a period of time to establish a baseline, if no baseline measurement is already done yet. From the baseline measurement, an average, standard deviation can be calculated. A boring range may be selected. Using mean and the standard deviation, for example the boring range is from ({overscore (d)}n−aσn) to ({overscore (d)}n+bσn), where a and b are some real numbers. Depending on the metric, the lower bound and the upper bound do not need to be symmetric. For example, the lower bound may be larger while upper bound is smaller, e.g. the boring range is ({overscore (d)}n−3σn) to ({overscore (d)}n+σn). The measurement period may be selected as 1 hour. Moreover, and as mentioned earlier, a boring threshold (as opposed to a boring range) may also be suitable for some metrics, in which case only one bound is defined, e.g. set a boring threshold as ({overscore (d)}n+σn), any value below the threshold is boring. The measurement period may vary depending on user preferences, but might usually be expect to be on the order of one hour FIGS. 6 and 7 illustrate an example in which the disclosed data reduction method is used to store, transmit, and report a certain metric, in this case the number of read operations that are performed at a given node as a function of time. FIG. 6 shows the raw data while FIG. 7 shows the data reported after using one of the data reduction methods of the current invention. The raw data, constituting 475 data points, is shown in FIG. 6, in which each time increment along the X-axis represents the number of read operations occurring within a five-second interval. A non-exponential running average and standard deviation are calculated every 95 time increments or so, and thus the mean and standard deviation are recomputed every 8 minutes or so, as can be seen in FIG. 7. Of course, the initial mean and standard deviation will be computed on the basis of some sort of historical data, which is not shown in the Figures for clarity. From this running mean and standard deviation calculation, a boring range is defined, which in this simple example represents the mean plus-or-minus one standard deviation. As noted earlier, boring values within the boring range are not reported. Thus, as shown in FIG. 7, when the boring values are removed, only 53 of the original 475 data points are deemed to be interesting and are reported, which represents approximately a nine-fold reduction in the amount of data that the monitoring system need deal with. Moreover, it can be seen that some of these interesting data points are either above the boring range (“large values”) or below the boring range (“small values”), and in this case only two of the 53 data points constitute such small values. Such large or small data points, when reported, may be treated differently be the system, as they may suggest different issues requiring different actions. However, it should be noted that this particular exemplary metric, read operations, is generally only interesting for monitoring purposes when large values occurs. Accordingly, in an alternative embodiment, one skilled in the art should note that only the upper bound for the metric (mean+one standard deviation) may be utilized for reporting purposes, which in effect would define a boring threshold as opposed to a boring range. If so configured, the number of interesting data points would be further reduced from 53 to 51, i.e., excluding the two small data values. In any event, whether defined by boring threshold or a boring range, the data that the system must handle is accordingly reduced. Still referring the example shown in FIGS. 6 and 7, the averages of the data change very little over time, although the standard deviations change a little more. In this case, the historic average and standard deviation may be used to define the boring range which would provide similar data reduction. FIG. 7 shows two border lines of the boring range, using only historic average and standard deviation. In this case, the number of interesting values is only about 12, rather than 53. For this example, using fixed average and boring range, data reduction would possibly be almost 40-fold. Another example is shown in FIGS. 8–9, File read operation data. The raw data in FIG. 8 shows the number of file read operation during each 15-minute interval. In FIG. 9, data close to the mean is replaced by the mean and the mean is updated along with the sampling. The interesting data reported is about one fifth of the raw data. FIGS. 10–13 present two more sets of examples showing the results of this embodiment of the current invention. In these two examples, rather than using predetermined fixed boring range based on historic data, the boring range are determined based on measured data. In these two examples, a method according to another embodiment of the current invention is employed such that the boring range is adjusted to match the moving trend in metrics. FIGS. 10–11 show CPU utilization over 22 hours, with computed standard deviation and running average. The FIG. 10 shows the original values. If a boring threshold is set based on historic data, as shown in FIG. 10, represented by a thick line, there are very few data points are boring. The data reduction is not substantial. A different data reduction method may be more suitable for this type of metrics. The FIG. 11 shows the original values when they differ from the running average by more than the standard deviation and shows the running average when it is closer. In FIG. 11, less than half the data points of FIG. 10 are shown, i.e. reported. In FIG. 11, 24 points are outside the boring range of the standard deviation, or are “interesting,” as denoted by diamonds. The moving average changed 12 times. Each time the moving average changes exceeding the predetermined threshold, the new average is reported to the monitor and will be used by the monitor in the future. Each new average is represented by a solid circle in FIG. 11. When a data point is within the boring range, it is replaced by the running average. These “boring” data point, which is replaced by the running average, is represented by a solid line. Since no data point is reported, no data point is displayed. In this example, there are 24 interesting data point and 12 changes of running average. Thus 36 values are reported instead of 90. The data reduction is about 3 times. FIG. 11 has less noise and all the important information in FIG. 10. So it actually highlights better what is important. In this example, when the running average change exceeds a defined criterion, then the running average is reported from the agent node to the console node, such that the console node may replace the old average with the new running average. The criteria of change may be based on the standard deviation calculated. In this example, one standard deviation is the criteria, i.e. when the current running average differs from the original running average (the average's initial value, which is also the average stored on the console) greater than the standard deviation, then the new running average is reported. The size of criteria is a trade-off between more raw data reporting (with poor average) or more averages reporting (with accurate average). FIGS. 12–13 show another work load data, with or without statistical data reduction. In this example, the single fixed threshold does not provide sufficient data reduction, as illustrated in FIG. 12. In FIG. 13, the running average is used, similar to the one used in the example in FIGS. 10–11. The average in FIG. 13 has changed 5 times during the monitoring period. After using the running average to update the average over time, the data reduction is about 4 times. Beside the data reduction, FIG. 13 also highlights the data trend that is not visible in the original data as shown in FIG. 12. Therefore, the method according to an embodiment of the current invention not only reduce the number of data points reported, but also reveals important information regarding the metric that is not obvious in the original raw data of the metric. Even though the amount of data reduction using this embodiment of the current invention may vary, depending on the type of metrics monitored, their distributions, errors tolerated, in all cases, the data reductions are substantial. It also provides a side benefit, i.e. highlighting the extraordinary events, which are most important to system performance monitoring and analysis. Another method to reduce the number of data collected, transferred between agent 302 and monitor console 420, is to reduce the number of metrics measured and monitored. When two or more metrics are highly correlated, then only the most important metric is measured, collected and transferred to the monitor console 420. The performance or activities of the other correlated metrics may be inferred from the reported metric. The correlation between two metrics can be measured by their correlation coefficient. A correlation coefficient is always between −1 and positive +1 (inclusive). When the correlation coefficient is +1, then the sequences have the identical shape or are completely related, as illustrated in FIG. 14. If the correlation coefficient is −1, then the two sequences are out of phase or moving in the completely opposite direction, as illustrated in FIG. 15. In both cases, when correlation coefficients are +1 or −1, the knowledge of one data sequence will provide complete knowledge regarding the trend of the other data sequence. Therefore, only the data for one sequence is needed for the performance monitoring or analysis for both data sequences. When the correlation coefficient equals to 0, the two data sequences are said to be independent, that is there is no relationship between the two sequences, as illustrated in FIG. 16. When the absolute value of correlation coefficient is between 0 and 1, there is some relationship between the two data sequence, as illustrated in FIG. 17, where the correlation coefficient is 0.5. When two metrics are highly correlated (for example the absolute value of the correlation coefficient is over some permissible threshold, e.g. 0.7), then it can be inferred that one will have a peak value when the other has a peak value. And when one metric reaches a trough then the other metric reaches a trough at the same time. Therefore knowing the movement of one metric, the movement of the other metric can be inferred. Based on the level of confidence c required, the amount of error e allowed, the sample size n can be determined, as described above. Accordingly, and on the basis of historic data, once the absolute value of the correlation coefficient is calculated and determined to be above the threshold, only the first metric will be sampled and reported, as described above. The second metric will not be sampled or reported. When the first metric has an “interesting” value and is reported, in one embodiment, the console may estimate the value of the second metric based on the correlation coefficient and the stored historic data. In another embodiment, the second metric is assumed to be the same as the first metric and the second metric is no longer monitored or analyzed. FIG. 18 shows one example where two metrics are closely correlated. As shown in FIG. 18, the file data operation and file write operation have a correlation coefficient of about 0.98. Therefore, peaks and troughs of the two metrics coincide. We can infer that allertable value of one will be allertable value of the other. Therefore, only one metrics is necessary to be monitored and measured. On the other hand, as shown in FIG. 19, the IP packets and web log hits have a correlation coefficient of 0.62. The trend of activity in IP packets therefore does not closely coincide with web log hits, and thus knowledge in one does not provide enough information to discern the performance of the other. If both metrics are necessary for performance monitoring, then both metrics have to be sampled and reported. Typically, the threshold of the absolute value of correlation coefficient is set at about 0.7. In some situations, higher threshold may be set, e.g. at 0.9 or 0.95. The higher the threshold, the more metrics need to be monitored. While illustrative embodiments of the invention have been illustrated and described, it will be appreciated that various changes can be made therein without departing from the spirit and scope of the invention as defined by the appended claims and their equivalents. |
|
abstract | A method is described for forming duplex layers including an interlayer and a corrosion resistant boundary layer on a nuclear fuel rod cladding tube for use in a water cooled nuclear reactor. The method includes forming, by thermal deposition or physical vapor deposition, on the exterior of a substrate, an inner interlayer with Mo, Ta, W or Nb or other particles, and forming, by thermal deposition or physical vapor deposition, on the interlayer, an outer corrosion resistant layer with particles selected from the group consisting of Cr, a Cr alloy, and combinations thereof. The interlayer prevents eutectic formation between the corrosion resistant layer and the substrate. |
|
claims | 1. A column assembly of a radionuclide generator, comprising:a column having an interior containing a medium for retaining a long-lived parent radionuclide that produces a relatively short-lived daughter radionuclide;an inlet port in fluid communication with the interior of the column; andan outlet port in fluid communication with the interior of the column;wherein the column assembly includes a vent opening that provides fluid access to the interior of the column via the outlet port, the vent opening oriented to face downwardly when the column is positioned below the outlet port during sterilization to prevent condensate from entering the vent opening from above. 2. The column assembly of claim 1, further comprising:an outlet cover that at least partially covers the outlet port and that at least partially defines the vent opening. 3. The column assembly of claim 2, wherein the outlet port includes a needle structure and the outlet cover includes a pierceable membrane that receives the needle structure of the outlet port. 4. The column assembly of claim 2, wherein the outlet cover includes a body portion and a removable cap. 5. The column assembly of claim 4, wherein the vent opening is defined as an annular space between the removable cap and the body portion. 6. The column assembly of claim 4, further comprising:a filter in the outlet cover. 7. The column assembly of claim 6, wherein the filter is bacteria retentive. 8. The column assembly of claim 6, wherein the filter is positioned at the vent opening. 9. The column assembly of claim 1, further comprising:a filter positioned between and in fluid communication with the outlet port and the column. 10. The column assembly of claim 1, wherein the inlet port is accessible from outside of a shielded package that receives the column assembly, when the column is inside of the shielded package. 11. The column assembly of claim 10, in combination with the shielded package. 12. The column assembly of claim 1, further comprising:a plug removably attached to the inlet port to block fluid communication to the inlet port from an atmosphere outside of the column assembly. 13. The column assembly of claim 1, wherein the medium includes alumina. 14. The column assembly of claim 1, in combination with the long-lived parent radionuclide and the relatively short-lived daughter radionuclide. 15. The combination of claim 14, wherein the long-lived parent radionuclide includes molybdenum-99 and the relatively short-lived daughter radionuclide includes technetium-99m. 16. A method of producing a terminally sterile column assembly of a radionuclide generator, comprising:providing a column assembly of a radionuclide generator that includes:a column having a long-lived parent radionuclide that produces a relatively short-lived daughter radionuclide;an inlet port in fluid communication with the column; andan outlet port in fluid communication with the column, the outlet port includes a vent opening that provides fluid access to the column;positioning the column assembly in an orientation with the vent opening facing downwardly to prevent condensate from entering the vent opening from above; andexposing the column assembly to steam for sterilization. 17. A column assembly of a radionuclide generator, comprising:a column and an outlet port, the column including a medium for retaining a long-lived parent radionuclide that produces a relatively short-lived daughter radionuclide, the outlet port in fluid communication with the column and covered with a vented outlet cover to provide a terminally sterilizable column assembly, the vented outlet cover having a vent opening that provides fluid access to the column and that prevents the ingress of gravity-driven liquid to produce a column assembly that consistently exhibits high yield. 18. A column assembly of a radionuclide generator, comprising:a column and an outlet port, the column including a medium for retaining a long-lived parent radionuclide that produces a relatively short-lived daughter radionuclide, the outlet port in fluid communication with the column and covered with a vented outlet cover to provide a terminally sterilizable column assembly, wherein means are provided to prevent the ingress of gravity-driven liquid to produce a column assembly that consistently exhibits high yield and that prevents migration of parent radionuclide away from the column. 19. A column assembly of a radionuclide generator, comprising:a column and an outlet port, the column including a medium for retaining a long-lived parent radionuclide that produces a relatively short-lived daughter radionuclide, the outlet port in fluid communication with the column and a means that allows for an exchange of steam while reliably preventing excess liquid from being introduced to portions of the column assembly during sterilization. 20. A column assembly of a radionuclide generator, comprising:a column and an outlet port, the column including a medium for retaining a long-lived parent radionuclide that produces a relatively short-lived daughter radionuclide, the outlet port in fluid communication with the column and covered with a vented outlet cover to provide a terminally sterilizable column assembly, the vented outlet cover having a vent opening that provides fluid access to the column and that prevents the ingress of gravity-driven liquid to produce a column assembly that prevents migration of parent radionuclide away from the column. |
|
claims | 1. An apparatus, comprising:a containment vessel having a first, fluid-containing region and a second, vacuum region, the second region housing a reactor pressure vessel of a nuclear reactor module, and the first region positioned above the second region;a control rod drive mechanism (CRDM) partially housed in the first region to control movement of control rods relative to the containment vessel;a fluid-tight barrier positioned around the CRDM, the fluid-tight barrier and the CRDM partitioning the first and second regions from each other wherein at least a portion of the CRDM is positioned in the second region and wherein one or more CRDM drive motors of the CRDM are thermally coupled with fluid in the first region;a fluid inlet positioned to direct the fluid into the first region, anda fluid outlet positioned to direct the fluid out of the first region. 2. The apparatus of claim 1, wherein the fluid inlet is positioned to direct air into the first region. 3. The apparatus of claim 2, further comprising an active device located outside the containment vessel to force the air outside the containment vessel through the fluid inlet into the first region. 4. The apparatus of claim 2, wherein the first region comprises air at a higher density than any air located in the second region. 5. The apparatus of claim 1, further comprising a heat exchanger positioned in thermal communication with the drive motors to remove heat generated by the drive motors from the containment vessel. 6. The apparatus of claim 1, wherein the fluid-tight barrier comprises an attachment section sealably coupled with the CRDM, and an expansion section coupled between the CRDM and the containment vessel to permit the attachment section to move with the CRDM relative to the containment vessel. 7. The apparatus of claim 6, wherein the CRDM is one of multiple CRDMs, and wherein individual CRDMs include a CRDM housing, and wherein the attachment section comprises a plate with an opening for each CRDM housing. 8. The apparatus of claim 6, wherein the expansion section comprises a bellows. 9. The apparatus of claim 6, wherein the attachment section is welded to the CRDM. 10. The apparatus of claim 1, wherein the at least the portion of the CRDM comprises one or more CRDM magnetic coils. 11. The apparatus of claim 1, wherein the fluid comprises air. |
|
052456447 | abstract | A spacer for a fuel assembly of a pressurized water reactor includes a first group of first webs standing on end and extending parallel to one another in a plane. Each of the first webs has longer sides and shorter sides. One of the longer sides of each of the first webs has a slit formed therein with a narrowed point and an impressed indentation. A second group of second webs stand on end and extend at right angles to the first webs. Each of the second webs has longer sides and shorter sides. Another of the longer sides of each of the second webs has a slit formed therein with a narrowed point and an impressed indentation. Each respective one of the first webs is connected to a respective one of the second webs to form a grid by inserting the webs into each other at the slits and locking the narrowed point of one web into place in the impressed indentation of another web with a plug-in connection. |
claims | 1. A cryogenic imaging system, comprising:a kinematic base that receives cartridges on a cryogenic base, with each cartridge carrying a specimen and further includes a positioning stage and a warm-cold interface between the positioning stage and the cryogenic base;a flexible thermal linkage between the cryogenic base and a refrigeration source to provide conductive cooling;a robotic loading and transfer system that is able to accept one or more cartridges and load the cartridges onto the cryogenic base;a source for generating an x-ray beam that irradiates the cartridges on the cryogenic base; anda detector for detecting the x-ray beam from the cartridges. 2. A system as claimed in claim 1, wherein the positioning stage positions the cryogenic base along three axes. 3. A system as claimed in claim 1, wherein the positioning stage rotates the cryogenic base. 4. A system as claimed in claim 1, wherein the warm-cold interface comprises a ball and groove configuration. 5. A system as claimed in claim 1, wherein the warm-cold interface comprises a ball and cone configuration. 6. A system as claimed in claim 1, wherein the flexible thermal linkage comprises one or more metal wires. 7. A system as claimed in claim 1, further comprising cryogenic shield surrounding the cryogenic base, the cryogenic shield being cooled by the refrigeration source. 8. A system as claimed in claim 7, wherein the flexible thermal linkage extends between the cryogenic base and the cryogenic shield. 9. A system as claimed in claim 7, wherein the cryogenic shield comprises beam ports through which the x-ray beam passes. 10. A system as claimed in claim 7, wherein the cryogenic shield comprises loading ports through which the robotic loading and transfer system transfers the cartridges. 11. A system as claimed in claim 1, wherein the robotic loading and transfer system comprises a shuttle holding multiple cartridges and a gripper for transferring cartridges between the shuttle and the cryogenic base. |
|
H00002097 | claims | 1. A pin discharge assembly for allowing elongate pins to be discharged from a source area having a source pressure, into a terminal area having a terminal pressure which is greater than the source pressure; comprising: a valve having a pin passageway therethrough the conveying elongate pins being discharged from the source area; the valve having flexible lip means having a duck bill shape and extending across the pin passageway and having a lip opening therethrough for conveying pins therethrough; the flexible lip means being collapsible to prevent backflow of gases into the source area; aspirator means connected downstream of the valve and having a pin passageway therethrough for communicating with the pin passageway in the valve to allow elongate pins to pass through the aspirator means; the aspirator means serving to intermittently produce an aspirator venturi pressure downstream of the flexible lip means which is less than the source pressure. a valve having an upstream end, downstream end, and a pin passageway therethrough for allowing elongate pins to be conveyed from the source area through the valve; the valve having flexible lip means extending across the pin passageway and having a lip opening for allowing pins to be conveyed therethrough; the lip means being operative to open positions by creating a pressure downstream of the lip means which is less then the source pressure upstream of the valve; aspirator means connected downstream of the valve and having a pin passageway therethrough for communicating with the pin passageway in the valve, to allow elongate pins to pass through the aspirator means; the aspirator means serving to produce an aspirator venturi pressure downstream of the flexible lip means which is less than the source pressure when pins are being conveyed or an aspirator static pressure which is greater than the source pressure when the discharge assembly is closed and awaiting discharge of another pin; a control valve means mounted downstream of the aspirator means to block the pin passageway and close the discharge assembly to prevent flow of aspiration gas through the aspirator means thereby automatically controlling said valve and aspirator means. 2. The pin discharge assembly of claim 1 further comprising a control valve mounted downstream of the aspirator means and adjustable between closed and open positions; the closed position blocking the pin passageway thereby causing an aspirator static pressure to exist downstream of the lip means to tightly seal the valve lip means and prevent backflow into the source area; the open position allowing elongated pins to pass through a pin passage port in the control valve and also cause fluid flow through the aspirator to reduce pressure downstream of the flexible lip means for easier passage of elongate pins therethrough. 3. The pin discharge assembly of claim 2 further comprising evacuation means operatively connected to the pin passageway downstream of the aspirator to remove fluid flowing from the source area through the pin passageway. 4. The pin discharge assembly of claim 2 further comprising a mechanical seal assembly having a pin passageway therethrough positioned upstream from the control valves to prevent leakage. 5. A pin discharge assembly for allowing elongate pins to be discharged from a source area having a source pressure, to a terminal area having a terminal pressure which is greater than the source pressure, comprising: |
summary | ||
claims | 1. A method of monitoring resistance welding quality of a nuclear fuel rod using a resistance welding apparatus, the resistance welding apparatus includinga tube electrode configured to contact a cladding tube of the nuclear fuel rod,a plug electrode configured to contact an end plug of the nuclear fuel rod,a current sensor for detecting welding current,a voltage sensor for detecting welding voltage,a welding force sensor for detecting welding force applied to the end plug, anda monitoring unit monitoring welding quality in real time by processing detected signals,the method comprising:providing the cladding tube to be inserted in the tube electrode and aligning the end plug on the cladding tube to come in contact with the cladding tube at one side of the end plug while an opposite side of the end plug is supported by the plug electrode;supplying alternating current (AC) electric power to the tube electrode and the plug electrode during a welding time such that the welding current flows through the tube electrode, the cladding tube, the end plug and the plug electrode, and moving the plug electrode such that the welding force is applied to the end plug, thereby welding the end plug to the cladding tube;detecting, at every predetermined sample cycle by the monitoring unit, the welding current using the current sensor, the welding voltage applied between the tube electrode and the plug electrode using the voltage sensor, and the welding force using the welding force sensor;calculating, by the monitoring unit, an entire effective current value, an entire effective voltage value, an entire average dynamic resistance, and an entire average heating value, whereinthe entire effective current value is an effective current of the welding current for the welding time,the entire effective voltage value is an effective voltage of the welding voltage for the welding time,the entire average dynamic resistance value is calculated using the entire effective current value and the entire effective voltage value, andthe entire average heating value is calculated using the entire effective current value, the entire effective voltage value and the welding time;comparing, by the monitoring unit, the entire effective current value, the entire effective voltage value, the entire average dynamic resistance value and the entire average heating value with predetermined reference values, respectively, wherein when the entire effective current value, the entire effective voltage value, the entire average dynamic resistance and the entire average heating value are not within a reference range of the predetermined reference values, respectively, the welding is determined to be defective;calculating, by the monitoring unit, an interval dynamic resistance value, a gradient value of instantaneous dynamic resistances, and a gradient value of instantaneous welding forces, whereinthe interval dynamic resistance value is calculated using an interval effective current and an interval effective voltage, the interval effective current value being an effective current of the welding current for each half cycle of the AC electric power, and the interval effective voltage value being an effective voltage of the welding voltage for said each half cycle of the AC electric power,the gradient value of the instantaneous dynamic resistances is an average gradient of the instantaneous dynamic resistances for said each half cycle of the AC electric power, each of the instantaneous dynamic resistances being calculated using the welding current and the welding voltage at said every predetermined sample cycle,the gradient value of the instantaneous welding forces is an average gradient of the instantaneous welding forces for said each half cycle of the AC electric power, each of the instantaneous welding forces being the welding force detected by the welding force sensor at said every predetermined sample cycle; andcomparing, by the monitoring unit, the interval dynamic resistance value, the gradient value of the instantaneous dynamic resistances and the gradient value of the instantaneous welding forces with a preset interval dynamic resistance value, a preset gradient value of instantaneous dynamic resistances, and a preset gradient value of instantaneous welding forces, respectively, wherein when the interval dynamic resistance value, the gradient value of the instantaneous dynamic resistances and the gradient value of the instantaneous welding forces are not within a predetermined range of the preset interval dynamic resistance value, the preset gradient value of instantaneous dynamic resistances, and the preset gradient value of instantaneous welding forces, respectively, the welding is determined to be defective. 2. The method of claim 1, wherein the gradient value of the instantaneous welding forces is the average gradient of the instantaneous welding forces at a first half cycle of the AC electric power. 3. The method of claim 1, wherein the performing of the welding includes forming a welding bead protruding in a ring shape at welded portion between the cladding tube and the end plug. 4. The method of claim 1, wherein the welding force sensor is a load cell, and the load cell includes a welding force indicator. 5. A method of monitoring resistance welding quality of a nuclear fuel rod using a resistance welding apparatus, the resistance welding apparatus includinga tube electrode configured to contact a cladding tube of the nuclear fuel rod,a plug electrode configured to contact an end plug of the nuclear fuel rod,a current sensor for detecting welding current,a voltage sensor for detecting welding voltage,a welding force sensor for detecting welding force applied to the end plug, anda monitoring unit monitoring welding quality in real time by processing detected signals,the method comprising:providing the cladding tube to be inserted in the tube electrode and aligning the end plug on the cladding tube to come in contact with the cladding tube at one side of the end plug while an opposite side of the end plug is supported by the plug electrode;supplying alternating current (AC) electric power to the tube electrode and the plug electrode during a welding time such that the welding current flows through the tube electrode, the cladding tube, the end plug and the plug electrode, and moving the plug electrode such that the welding force is applied to the end plug, thereby welding the end plug to the cladding tube;detecting, at every predetermined sample cycle by the monitoring unit, the welding current using the current sensor, the welding voltage applied between the tube electrode and the plug electrode using the voltage sensor, and the welding force using the welding force sensor;calculating, by the monitoring unit, an entire effective current value, an entire effective voltage value, an entire average dynamic resistance, and an entire average heating value, whereinthe entire effective current value is an effective current of the welding current for the welding time,the entire effective voltage value is an effective voltage of the welding voltage for the welding time,the entire average dynamic resistance value is calculated using the entire effective current value and the entire effective voltage value, andthe entire average heating value is calculated using the entire effective current value, the entire effective voltage value and the welding time;comparing, by the monitoring unit, the entire effective current value, the entire effective voltage value, the entire average dynamic resistance value and the entire average heating value with predetermined reference values, respectively, wherein when the entire effective current value, the entire effective voltage value, the entire average dynamic resistance and the entire average heating value are not within a reference range of the predetermined reference values, respectively, the welding is determined to be defective;calculating, by the monitoring unit, an interval dynamic resistance value, a gradient value of instantaneous dynamic resistances, and a gradient value of instantaneous welding forces, whereinthe interval dynamic resistance value is calculated using an interval effective current and an interval effective voltage, the interval effective current value being an effective current of the welding current for each half cycle of the AC electric power, and the interval effective voltage value being an effective voltage of the welding voltage for said each half cycle of the AC electric power,the gradient value of the instantaneous dynamic resistances is an average gradient of the instantaneous dynamic resistances for said each half cycle of the AC electric power, each of the instantaneous dynamic resistances being calculated using the welding current and the welding voltage at said every predetermined sample cycle,the gradient value of the instantaneous welding forces is an average gradient of the instantaneous welding forces for said each half cycle of the AC electric power, each of the instantaneous welding forces being the welding force detected by the welding force sensor at said every predetermined sample cycle; andcomparing, by the monitoring unit, the interval dynamic resistance value, the gradient value of the instantaneous dynamic resistances and the gradient value of the instantaneous welding forces with a preset interval dynamic resistance value, a preset gradient value of instantaneous dynamic resistances, and a preset gradient value of instantaneous welding forces, respectively, whereinwhen the interval dynamic resistance value is within a predetermined range of the preset interval dynamic resistance value, a first quantified value is added to a total quantified value,when the gradient value of the instantaneous dynamic resistances is within a predetermined range of the preset gradient value of instantaneous dynamic resistances, a second quantified value is added to the total quantified value,when the gradient value of the instantaneous welding forces is within a predetermined range of the preset gradient value of instantaneous welding forces, a third quantified value is added to the total quantified value, andwhen the total quantified value is greater than a predetermined total quantified value, the welding is determined to be normal. 6. The method of claim 5, wherein when the total of the first, second and third quantified values is less than the predetermined total quantified value, the welding is defective or warning depending on a level of the total. 7. The method of claim 5, wherein the first quantified value, the second quantified value and the third quantified value are weighted. |
|
055240321 | abstract | A cladding tube having a cross-section and including (1) a zirconium alloy outer circumferential substrate having an inner surface and having one or more alloying elements, (2) a zirconium barrier layer bonded to the inner surface of the outer circumferential substrate and being alloyed with the one or more alloying elements, and (3) a zirconium alloy inner circumferential liner bonded to the inner surface of the zirconium barrier layer. The barrier layer will have a concentration profile including a diffusion layer extending from the barrier layer's inner surface (facing nuclear fuel) to the barrier layer's interior (between the barrier layer's inner and outer surfaces). At the interior edge of the diffusion layer, there will be substantially no alloying elements. At the outer edge of the diffusion layer (the barrier layer's inner surface), the maximum concentration of alloying elements will occur. |
041750039 | summary | BACKGROUND OF THE INVENTION The invention described herein relates to nuclear reactor fuel assemblies and more particularly to an improved grid used in holding fuel rods and control rod guide tubes in spaced relationship with each other in a fuel assembly. Many different designs of grids are used for holding fuel rods and control rod guide tubes in spaced relationship with each other in a nuclear reactor fuel assembly. One well known design disclosed in U.S. Pat. No. Re. 28,079 includes interleaved grid straps which form multiple cells arranged to contain the fuel rods and guide tubes. The fuel rod in each cell is supported by springs on two adjacent straps which project inwardly and contact two sides of the fuel rods. Each of the other two adjacent straps contain a pair of vertically spaced dimples which contact the other two sides of the fuel rod thus providing a six point support to the rod. This design provides many benefits, such as resiliently supporting fuel rods in each egg-crate cell and acting to accurately space the fuel rods while minimizing fuel rod vibration and permitting fuel rod thermal expansion during reactor operation. The disadvantages however are that the dimples occasionally are torn from the metal surface during the time a fuel rod is being loaded into the grid and since the fuel rod is then not properly supported, the grid must be removed from the fuel assembly being assembled. Further, special dies are required to punch out the springs and dimples with a high degree of precision and accuracy to assure that coolant flow distribution along the fuel rods and through the grids will be uniform and constant. Manufacturing labor costs are therefore relatively high. Also, since the grids usually are made of Inconel, which work hardens very quickly, cracks occasionally appear at the point where the dimples are punched outwardly from the stock material. Since extremely high standards and tolerances are set for the manufacture and quality of grid components, small defects which would not affect reactor performance often require scrapping of the grid material. SUMMARY OF THE INVENTION The above described disadvantages are eliminated in accordance with this invention by providing a grid of interleaved straps arranged to form multiple cells designed to contain fuel rods. Each of two adjacent straps contain springs having two contact points which project inwardly from the strap walls while the other two adjacent straps which cooperate with the first two to form a cell, contain springs, each having one contact point, which likewise project inwardly from the strap walls into the cell. As a fuel rod is drawn into the cell, it rides up contoured walls on the spring surface and into position to be supported by a total of six contacts on the four springs. An object of the invention, therefore, is to provide an improved grid for a nuclear reactor fuel assembly which will impart improved support to fuel rods and minimize fretting of fuel assembly components. Another object of the invention is to provide an improved grid for a nuclear reactor fuel assembly which can be manufactured at a cost less than current designs while reducing die maintenance costs and inspection operations. |
summary | ||
summary | ||
description | The present application is the US national stage of PCT/CN2010/073041 filed on May 21, 2010, which claims the priority of the Chinese patent application No. 200910107548.5 filed on May 27, 2009, which application is incorporated herein by reference. The invention relates to a method for joint configuration design of fuel elements/assemblies, and more particularly relates to a method for joint configuration design of nuclear power plant fuel elements/assemblies so as to improve the fuel utilization. Nuclear power is a new energy which has been paid more and more attention since the first nuclear power plant was built in 1950s. According to statistics, about 441 nuclear power plants have been built in more than 30 countries and regions in the world by the end of 2002, and the generation capacity is about 360 million kilowatts. Now about 40 nuclear power plants are being built and about 60 nuclear power plants are to be built. After completion, the generation capacity will be up to 500 million kilowatts which will take up about 20% of the world's total generating capacity at that time. Uranium or plutonium are generally utilized as nuclear fuel in the nuclear reactor, wherein, the fission chain reaction is conducted in a controlled manner and then fission energy is produced continuously from a special atomic boiler for doing work. Now a research for joint configuration of nuclear fuel elements/assemblies among the different reactor types of the nuclear power plant is being conducted. As the shape of fuel elements/assemblies varies as the reactor type, during the joint configuration of nuclear fuel elements/assemblies among different reactor types, the fuel elements/assemblies will experience disintegration and recombinant. While among several units of the same reactor type with interchangeable fuel elements/assemblies, the configuration design of the nuclear fuel elements/assemblies in reactor core is relatively independent and lack of unified coordinate optimization. Under the condition where the configuration design of the nuclear fuel elements/assemblies in reactor core is relatively independent, the fuel elements/assemblies loaded in the first reactor core (it is the core of the first cycle for the reactor type refueling when shutdown) generally are all new fuel elements/assemblies. There three enrichments for the new fuel assemblies are generally used. To take the first reactor core of Daya Bay Nuclear Power Station for example, three enrichments by mass of uranium 235 for the new fuel assemblies are 1.8%, 2.4% and 3.1% respectively. A method for optimizing the arrangement of the enrichments and quantity of the first reactor core according to the reactivity distribution in the equilibrium cycle is provided in a Chinese published application CN 200510071705.3 of Westinghouse Company, so as to save uranium resources and reach the equilibrium cycle more quickly. The first reactor core configuration design is entirely independent and all the fuel assemblies loaded are still new, except that the range of the enrichments is wider, the types of the new fuel assemblies are more and a low-leakage loading patterns is utilized. In the method provided in a Japanese published application JP 61-66988A of Toshiba Company, the spent fuel assemblies that can not be used by the operating units any longer are loaded in the first reactor core of the boiling water reactor to replace the new fuel assemblies with natural enrichment, so as to save the natural uranium, manufacturing and the spent fuel disposal cost of the replaced new fuel assemblies. However, this method has serious limitations. For example, in the first reactor core of the pressurized water reactor, the enrichments of the new fuel assemblies in use are generally much higher than that of nature uranium (for example, the smallest enrichment of the new fuel assemblies used in the first reactor core of Daya Bay Nuclear Power Station is 1.8% that is much higher than that of natural uranium with value of 0.71%), and there is no suitable alternative in the first core for the spent fuel assemblies that can not be used any longer by the operating units, resulting in small scope of application and less practical value to other reactor types, and just saving the cost of the new fuel assemblies with natural enrichment for the boiling water reactor which has small effect compared with the cost of the first core. Moreover, the first reactor core configuration design is still completely independent, and just the spent fuel assemblies that can not be used by the operating units any longer is loaded, such that there is no interaction and effect on the reactor core configuration design of the units that have been running An objective of this invention is aimed to the disadvantage that, in the prior art, among several units of the same reactor type with interchangeable fuel elements/assemblies, the configuration design of the nuclear fuel elements/assemblies in reactor core is relatively independent and lack of unified coordinate optimization, and the fuel burnup is low, and then to provide a method for joint configuration design of nuclear power plant fuel elements/assemblies. A technical solution utilized in this invention to solve its technical problem is providing a method for joint configuration design of nuclear power plant fuel elements/assemblies, comprising steps implemented in fuel elements/assemblies joint configuration design conducted between at least one operating unit and at least one new starting unit with the first core: (S1) for at least one operating unit, based on an equilibrium cycle or transition cycle reactor core configuration design, adding at least one new fuel element/assembly to the at least one operating unit, wherein, for a reactor type refueling when non-shutdown, adding at least one new fuel element/assembly means adding more new fuel elements/assemblies in a period of time; (S2) after running a fuel cycle, and on the basis of the added new fuel elements/assemblies loaded in step (S1), obtaining more spent fuel elements/assemblies having burned once from the at least one operating unit than that obtained in the reactor core configuration design of the equilibrium cycle or original transition cycle, and keeping said spent fuel elements/assemblies in reserve; wherein, with respect to the reactor type refueling when shutdown, after running one fuel cycle, based on the added new fuel elements/assemblies loaded in step (S1), more spent fuel elements/assemblies having burned once are obtained from the at least one operating unit than that are obtained in the reactor core configuration design of the equilibrium cycle or original transition cycle for reserve; with respect to the reactor type refueling when non-shutdown, after the at least one operating unit has been running for a period of time, based on the added new fuel elements/assemblies loaded in step (S1), more spent fuel elements/assemblies having burned partially are obtained than that are obtained in the reactor core configuration design of the equilibrium cycle or original transition cycle for reserve. (S3) for at least one new starting unit with the first core, arranging a scheduled number of new fuel elements/assemblies and the spent fuel elements/assemblies in reserve having burned once which are obtained from step (S2) in the first reactor cores of at least one new starting unit, wherein, for at least one new starting unit with the first core, with respect to the reactor type refueling when shutdown, the scheduled number of new fuel elements/assemblies and the spent fuel elements/assemblies in reserve having burned once which are obtained from step (S2) are arranged in the first reactor cores of at least one new starting unit; with respect to the reactor type refueling when non-shutdown, the scheduled number of new fuel elements/assemblies and the spent fuel elements/assemblies in reserve having burned partially which are obtained from step (S2) are arranged in the first reactor cores of at least one new starting unit. The spent fuel elements/assemblies having burned once and the spent fuel elements/assemblies having burned twice described herein are defined by the roughly reactivity after the fuel elements/assemblies have burned once and twice respectively, rather than by the real loaded times. For the reactor core comprising the new fuel elements/assemblies, the spent fuel elements/assemblies having burned once, the spent fuel elements/assemblies having burned twice and the spent fuel elements/assemblies having burned more times, the method for obtaining and using the spent fuel elements/assemblies in reserve having burned once which is mentioned in steps (S1), (S2) and (S3) also comprises the method for obtaining and using the spent fuel elements/assemblies having burned once, the spent fuel elements/assemblies having burned twice and the spent fuel elements/assemblies having burned more times in such reactor core. The method for joint configuration design of nuclear power plant fuel elements/assemblies of the present invention, between the steps (S2) and (S3) further comprises storage of the spent fuel elements/assemblies in reserve having burned once which are obtained from step (S2) in the spent fuel pool or spent fuel storage container, so as to satisfy the requirements of transportation safely after storage, that is storing the spent fuel elements/assemblies having burned once and the spent fuel elements/assemblies having burned twice in step (S2) (with respect to the reactor type refueling when shutdown), and the spent fuel elements/assemblies having burned partially (with respect to the reactor type refueling when non-shutdown), once the requirements for transportation comprising the criticality safety, radiation shielding, cooling and so on are satisfied, the spent fuel elements/assemblies in reserve to be used are transported from the operating units to the new starting units with the first cores by the transporting container of spent fuel. In the method for joint configuration design of nuclear power plant fuel elements/assemblies of the present invention, the arrangement of the first reactor core of the at least one new starting unit is based on the refueling mode in which the at least one operating units are added with the new fuel elements/assemblies before or after. In the method for joint configuration design of nuclear power plant fuel elements/assemblies of the present invention, the spent fuel elements/assemblies having burned once and the new fuel elements/assemblies are arranged alternatively in the reactor core of the at least one operating unit or in the first reactor core of the new starting unit, such that the new fuel elements/assemblies are arranged around the spent fuel elements/assemblies having burned once. Such alternative arrangement can improve the burnup of the spent fuel elements/assemblies having burned once in the present cycle, improve the discharge burnup, and improve the fuel efficiency. In the method for joint configuration design of nuclear power plant fuel elements/assemblies of the present invention, the step (S3) also comprises arranging a scheduled number of spent fuel elements/assemblies having burned twice taken from the at least one operating unit into the first reactor core of the new starting unit. In the method for joint configuration design of nuclear power plant fuel elements/assemblies of the present invention, in step (S1), for at least one operating unit, 1˜29 new fuel assemblies are added to the at least one operating unit, based on the equilibrium cycle or transition cycle reactor core configuration design; wherein, with respect to the reactor core with 157 fuel assemblies, 1˜29 new fuel assemblies are added to the at least one operating unit, such that number of the new fuel assemblies is up to 101; the reactor core with more fuel assemblies can be added with more new fuel assemblies; with respect to the reactor type refueling when non-shutdown, the addition means loading more new fuel assemblies during a period of time. In the method for joint configuration design of nuclear power plant fuel elements/assemblies of the present invention, 1, 4, 5, 8, 9, 12, 13, 16, 17, 20, 21, 24, 25, 28 or 29 new fuel assemblies are added to the at least one operating unit. The method of adding fuel elements/assemblies is applicable to the reactor core having the central fuel assembly and showing a quarter-symmetry. The added quantity varies correspondingly with respect to the reactor core showing other symmetry or in other refueling modes. In the method for joint configuration design of nuclear power plant fuel elements/assemblies of the present invention, in step (S1), the reactor cores of the at least one operating unit have 157 fuel assemblies; In step (S3), the first reactor cores of the at least one new starting unit comprise 80˜100 new fuel assemblies, 30˜49 spent fuel assemblies having burned once and the rest spent fuel assemblies having burned twice. In the method for joint configuration design of nuclear power plant fuel elements/assemblies of the present invention, there are several operating units which comprise long-term refueling and short-term refueling units; and there are steps implemented in the joint configuration of fuel assemblies conducted among the operating units: (B1) based on the reactor cores configuration design of the long-term equilibrium cycle or transition cycle, the spent fuel assemblies having burned once which are arranged in the peripheral places of the loading pattern are replaced by the spent fuel assemblies having burned twice, so as to obtain excess spent fuel assemblies having burned once for reserve; (B2) the excess spent fuel assemblies having burned once for reserve and the spent fuel assemblies in the long-term operating units are arranged in the short-term operating units. In the method for joint configuration design of nuclear power plant fuel elements/assemblies of the present invention, the long-term operating units described herein are the units refueling every 18 months or every 24 months; the short-term operating units described herein are the units refueling quarter or third of the fuel assemblies every year. In the method for joint configuration design of nuclear power plant fuel elements/assemblies of the present invention, the new fuel assemblies are with relatively high enrichments equal to that of the new fuel assemblies used in the equilibrium cycle. By implementing the method for joint configuration design of nuclear power plant fuel elements/assemblies of the present invention, there are following advantageous effects: several operating units with interchangeable fuel elements/assemblies are joined for nuclear fuel elements/assemblies configuration, such that the operating units and the new starting units with the first cores, and the operating units in different refueling modes can interact, coordinate optimization, and complement each other respectively; all the new fuel elements/assemblies are with relatively high enrichments same or equivalent to that of the new fuel elements/assemblies used in the equilibrium cycle; in the first reactor core of the new starting units, a part of spent fuel elements/assemblies have burned a certain number of cycles, such as one, two or more cycles, while burned for a period time with respect to the reactor type refueling when non-shutdown, so as to satisfy the reactivity requirements for constructing the first reactor cores of the new starting units, resulting in that a variety of desired refueling modes can be implemented in the first reactor cores for the reactor type refueling when shutdown; furthermore, the ability of units to resist risks (a number of fuel elements/assemblies are damaged) can be improved, the method is suitable for both the reactor type refueling when shutdown and non-shutdown, and for (pressurized-water reactor) (PWR), boiling water reactor (BWR), heavy water reactor (HWR), and high temperature reactor (HTR). The technical solution of the present invention is about a design for nuclear fuel elements/assemblies configuration conducted jointly by several units with interchangeable fuel elements/assemblies. The first core of the new starting units may totally use the fuel elements/assemblies with relatively high enrichments which are identical or equivalent to the equilibrium cycle enrichments. Wherein, a part of new fuel elements/assemblies with relatively high enrichments may be put into one or more operating units for combustion, such that the new fuel elements/assemblies loaded in the operating units is much more than when such method is not implemented, and then considerable spent fuel elements/assemblies having burned once are obtained. Meanwhile, considerable spent fuel elements/assemblies having burned twice may be provided by the operating units additionally (taking the reactor core with 157 fuel assemblies refueling every 18 months for example). From the considerable obtained spent fuel elements/assemblies having burned once and spent fuel elements/assemblies having burned twice, the spent fuel elements/assemblies with the desired quantity and burnup may be selected to construct the first reactor cores of the new starting units together with new fuel elements/assemblies with relatively high enrichments. After the method for joint configuration design of nuclear fuel elements/assemblies has been implemented, the average discharge burnup of the operating units added with new fuel elements/assemblies and the first reactor cores of the new starting units (if the arrangement mode is utilized in which the operating units have been added with new fuel elements/assemblies) is higher than that of the equilibrium cycle (generally, the average discharge burnup of the first reactor cores of the new starting unit is less than 50% of that of the equilibrium cycle). Specific steps are as follows: (S1) for at least one operating unit, based on the reactor core configuration design of the equilibrium cycle or transition cycle, at least one new fuel element/assembly is added to at least one operating unit, such as 1˜29 new fuel assemblies; (S2) after running a fuel cycle, and on basis of the new fuel elements/assemblies added in step (S1), more spent fuel elements/assemblies having burned once which are obtained from the at least one operating unit than that are obtained in the reactor core configuration design of the equilibrium cycle or transition cycle, and said spent fuel elements/assemblies are kept in reserve; the spent fuel elements/assemblies in reserve having burned once which are obtained from step (S2) are stored in the spent fuel pool; once the requirements for transporting comprising the criticality safety, radiation shielding, cooling and so on are satisfied, the spent fuel elements/assemblies in reserve to be used may be transported from the operating units to new starting units with the first cores by the transporting containers of spent fuel. (S3) for at least one new starting unit with the first core, a scheduled number of new fuel elements/assemblies, the spent fuel elements/assemblies in reserve having burned once which are obtained from step (S2), and a scheduled number of the spent fuel elements/assemblies having burned twice are arranged in the first reactor cores of at least one new starting unit. In the reactor core configuration design of the present invention, the reactor core with 157 fuel assemblies of pressurized-water reactor (PWR) nuclear power plant refueling every 18 months is taken as an example for illustration, wherein, the reactor core comprises the new fuel elements/assemblies, the spent fuel elements/assemblies having burned once and the spent fuel elements/assemblies having burned twice. The spent fuel elements/assemblies having burned once and the spent fuel elements/assemblies having burned twice described herein are defined by the roughly reactivity after the fuel elements/assemblies have burned once and twice respectively, rather than by the real loaded times. For the reactor core comprising the new fuel elements/assemblies, the spent fuel elements/assemblies having burned once, the spent fuel elements/assemblies having burned twice and the spent fuel elements/assemblies having burned more times, the method for obtaining and using the spent fuel elements/assemblies in reserve having burned once which is mentioned in steps (S1), (S2) and (S3) also comprises the method for obtaining and using the spent fuel elements/assemblies having burned once, the spent fuel elements/assemblies having burned twice and the spent fuel elements/assemblies having burned more times in such reactor core. There is no restriction for the refueling mode of the operating units involved in the method for the joint configuration of nuclear power plant fuel elements/assemblies according to the present invention, and a variety of refueling mode may be implemented directly in the first cores of the new starting units. The method is suitable for the reactor types both with refueling when shutdown and with refueling when non-shutdown, and also for PWR, BWR, DUR, and HTR. To take the PWR nuclear power plant having the reactor with 157 fuel elements/assemblies and refueling when shutdown for example, exemplary embodiments about the fuel elements/assemblies joint configuration of one operating unit and one same type new starting unit with the first core will be illustrated as followed. FIG. 13 is the schematic diagram of joint configuration of nuclear power plant fuel assemblies with only one new starting unit with the first core. The arrows in the figure just indicate the transfer of the spent fuel elements/assemblies having burned once and the spent fuel elements/assemblies having burned twice. While the new fuel elements/assemblies are purchased from suppliers. The twelfth cycle of operating unit NO. 1, the fourteenth cycle of operating unit NO. 1 and the second cycle of new starting unit NO. 1 shown in FIG. 13 are all in the known equilibrium cycle refueling every 18 months in which 72 new fuel assemblies loaded. The other two cycles that is the thirteenth cycle of operating unit NO. 1 and the first cycle of new starting unit NO. 1 are both in the mode refueling every 18 months in which 88 new fuel assemblies loaded. Specially, as compared to 72 new fuel assemblies with an enrichment of 4.45% needed in the equilibrium cycle, additional 16 new fuel assemblies with an enrichment of 4.45% are added in the thirteenth cycle of operating unit NO. 1, such that the new fuel assemblies total up to 88. And the first cycle of new starting unit NO. 1 (the first reactor core) can obtain 41 spent fuel assemblies having burned once and 28 spent fuel assemblies having burned twice from the operating unit NO. 1. FIG. 1 shows the equilibrium cycle of the operating units refueling every 18 months. 72 new fuel assemblies with an enrichment of 4.45% are added to each cycle. The reactor core comprises 72 new fuel assemblies, 69 spent fuel assemblies having burned once and 16 spent fuel assemblies having burned twice. The reactor core configuration design is in symmetry of quarter rotation, and what shown in the figure is one quarter of the reactor core. Because of the single fuel assembly at the center of the reactor core, the quantity of the new fuel assemblies is not equal to that of the spent fuel assemblies having burned once. At the end of the twelfth cycle of operating unit NO. 1, 72 spent fuel assemblies having burned once and 69 spent fuel assemblies having burned twice are obtained, as shown in FIG. 15. As shown in FIG. 13, 16 new fuel assemblies with an enrichment of 4.45% are added to the thirteenth cycle of operating unit NO. 1. Such that the number of the new fuel assemblies in the thirteenth cycle of operating unit NO. 1 is increased to 88. With the fuel cycle length guaranteed, the number of the needed spent fuel assemblies having burned once as shown in FIG. 9 is reduced to 41, and the number of the spent fuel assemblies having burned twice is increased to 28. As a result, the reactor core now comprises 88 new fuel assemblies, 41 spent fuel assemblies having burned once and 28 spent fuel assemblies having burned twice. As shown in FIG. 15, after the burned assemblies discharged at the end of the twelfth cycle of operating unit NO. 1 are reloaded to construct the thirteenth cycle of operating unit NO. 1, 31 (72−41=31) spent fuel assemblies having burned once and 41 (69−28=41) spent fuel assemblies having burned twice are remained in reserve. As shown in FIG. 16, after the thirteenth cycle of operating unit NO. 1 with the number of the new fuel assemblies increased to 88 has passed one fuel cycle, 88 spent fuel assemblies having burned once and 41 spent fuel assemblies having burned twice can be obtained. The fourteenth cycle of operating unit NO. 1 returns into the loading mode of equilibrium circle that is shown in FIG. 1 where there are 72 new fuel assemblies, 69 spent fuel assemblies having burned once and 16 spent fuel assemblies having burned twice, so 69 spent fuel assemblies having burned once and 16 spent fuel assemblies having burned twice are required. As shown in FIG. 16, after the construction of the fourteenth cycle of operating unit NO. 1 is completed, 19 (88−69=19) spent fuel assemblies having burned once and 25 (41−16=25) spent fuel assemblies having burned twice are remained in reserve. As shown in FIG. 17, by only one flexibility cycle of the thirteenth cycle of operating unit NO. 1 in which 16 new fuel assemblies are added, 50 (31+19=50) spent fuel assemblies having burned once and 66 (41+25=66) spent fuel assemblies having burned twice can be obtained for reserve. The first reactor core of the new starting unit is to be constructed according to the loading mode with 88 new fuel assemblies as shown in FIG. 9 in which there are 88 new fuel assemblies, 41 spent fuel assemblies having burned once and 28 spent fuel assemblies having burned twice. So the first reactor core of the new starting unit requires 88 new fuel assemblies, 41 spent fuel assemblies having burned once and 28 spent fuel assemblies having burned twice. As shown in FIG. 17, 41 spent fuel assemblies having burned once and 28 spent fuel assemblies having burned twice are respectively selected from 50 spent fuel assemblies having burned once and 66 spent fuel assemblies having burned twice obtained for reserve. Together with 88 new fuel assemblies, they are used to construct the first reactor core of the new starting unit. Or as shown in FIG. 2 and FIG. 3, 80 new fuel assemblies, 49 spent fuel assemblies having burned once and 28 spent fuel assemblies having burned twice are loaded to construct the first reactor core of the new starting unit. Or as shown in FIG. 4 and FIG. 5, 81 new fuel assemblies, 48 spent fuel assemblies having burned once and 28 spent fuel assemblies having burned twice are loaded. Or as shown in FIG. 6 and FIG. 7, 84 new fuel assemblies, 45 spent fuel assemblies having burned once and 28 spent fuel assemblies having burned twice are loaded. Or as shown in FIG. 8, 85 new fuel assemblies, 44 spent fuel assemblies having burned once and 28 spent fuel assemblies having burned twice are loaded. Or as shown in FIG. 10, 89 new fuel assemblies, 44 spent fuel assemblies having burned once and 28 spent fuel assemblies having burned twice are loaded. Or as shown in FIG. 11, 92 new fuel assemblies, 37 spent fuel assemblies having burned once and 28 spent fuel assemblies having burned twice are loaded. Or as shown in FIG. 12, 93 new fuel assemblies, 36 spent fuel assemblies having burned once and 28 spent fuel assemblies having burned twice are loaded. As show in FIG. 2˜12, with the increase of new fuel assemblies loaded, the arrangement of the new fuel assemblies and the spent fuel assemblies having burned once at the central region of the reactor core tends to be checkerboard. The checkerboard arrangement as known can improve the burnup of the spent fuel assemblies having burned once in the present cycle, improve discharge burnup and improve fuel efficiency. In the loading mode of the second cycle of new starting unit NO. 1, there are 72 new fuel assemblies refueling every 18 months, which is identical to that in the fourteenth cycle of operating unit NO. 1, so the illustration will not be repeated again. The quantity of the new fuel assemblies additionally added and the loading mode of the first reactor cores of the new starting units may be adjusted based on the quantity of the first cores of new starting units and of the flexibility cycles ready to run in the operating units which are involved in the method for joint configuration design of nuclear power plant fuel elements/assemblies. For the reactor core described above with 18-month refueling and low leakage, 80, 81, 84, 85, 88, 89, 92 or 93 new fuel assemblies may be added, referring to the loading patterns as shown in FIG. 2˜FIG. 12 respectively. As the present inventing is not limited to the low leakage loading mode of in-in-out described above, in-out-in, high leakage (a part of new assemblies being placed in the peripheral place) and other loading mode are also applicable. So more reactor core loading modes may be utilized and more new fuel assemblies with relatively high enrichments may be added to the operating units. The loading patterns described above are just used for illustrating rather than limiting. The in-in-out loading mode mentioned above is the fuel elements/assemblies loading mode in which new fuel elements/assemblies are placed in the central places during the first burning cycle of the fuel elements/assemblies, in the central places during the second burning cycle of the fuel elements/assemblies, and in the peripheral places during the third burning cycle of the fuel elements/assemblies. While the in-out-in loading mode (able to lower the temperature difference at the outlet of the reactor core) is the fuel elements/assemblies loading mode in which new fuel elements/assemblies are placed in the central places during the first burning cycle of the fuel elements/assemblies, in the peripheral places during the second burning cycle of the fuel elements/assemblies, and in the central places again during the third burning cycle of the fuel elements/assemblies. The spent fuel elements/assemblies having burned are different from the new fuel elements/assemblies. The spent fuel elements/assemblies themselves have strong radiation and decay heat, and FIG. 18 shows the decay heat of a single fuel element/assembly as a function of cooling time. The decay heat decreases rapidly in the early phase, and slowly in the later phase. In the limiting case with an enrichment of 4.45% and a burnup of 55000 MWd/tU, the decay heat after one-year cooling is less than 6.1 kW, while the actual average decay heat after one-year cooling of the spent fuel elements/assemblies is less than 4 kW. If the spent fuel elements/assemblies are to be transported from one unit to another, a transporting container of spent fuel which satisfies the requirements is needed, wherein, the longer cooling time, the lower the requirements. There are a large number of mature commercial large transporting containers of spent fuel which may be used for transporting the spent fuel elements/assemblies having been cooled for only one year. The known transporting container of spent fuel which can transport the spent PWR fuel elements/assemblies with more than 5 kW decay heat are shown in the following table 1. TABLE 1Technical RequirementsMaximumAverageheat loadTheof theNumber ofMaximumSpent FuelLoadedHeatAssembliesTotal ContainerSpent FuelLoad(kW perWeightOwnerModelAssemblies(kW)one)(t)GEIF-3000 768-709.7-10 60.7BNGNTL-14 5 459 85 LehrerLK-80121008.3100 TNTN-12/2(A/B)1293/707.8/5.8102/104NACNLI-1010 707 97 TNTN-17/2 7 436.181 PNTLEXCELLOX-3A 5 306 72 PNTLEXCELLOX-4 7 405.792 OCLHZ-75T 7 405.782 TNTN-1312 645.3105 GNSCASTOR-S1 6 305 79-82 During the loading, unloading and transporting of the spent fuel elements/assemblies, it is necessary to ensure the critical security, cooling, shielding and isolation from the oxygen so as to avoid excessive oxidation of fuel cladding. By implementing the joint configuration of nuclear power plant fuel elements/assemblies, the average discharge burnup of the operating units and the first reactor cores of the new starting units is 17% higher than that of the 18-month equilibrium cycle in the prior art (generally, the average discharge burnup of the first circle of the new starting unit is less than 50% of that of the equilibrium cycle). Such that the first reactor core of every new starting unit in the method for joint configuration design of nuclear power plant fuel elements/assemblies may save about 200,000 pounds of natural uranium, about 32 tons SWU (separative work units) of separative work, about fabrications of 63 fuel assemblies, and about post-processes of 63 spent fuel elements/assemblies, resulting in the total savings of about $69.63 million which take up 23.4% of the total fuel costs of the first core. The cost savings may vary as the model, the reactor type, the joint configuration of the nuclear power plant fuel elements/assemblies and the price of natural uranium and separation work. With the new starting units with the first cores adding continuously, the joint configuration of the nuclear power plant fuel elements/assemblies can be implemented continuously (as shown in FIG. 14). The arrows in the figure just indicate the transfer of the spent fuel elements/assemblies having burned once, and the spent fuel elements/assemblies having burned twice are from what accumulated in the long-term running process of the operating units, and the new fuel elements/assemblies are purchased from fuel suppliers. For units refueling when shutdown, the operating units implementing the joint configuration of the nuclear power plant fuel elements/assemblies may utilize 18-month refueling mode, 12-month refueling mode with quarter of fuel assemblies changed, 12-month refueling mode with third of fuel assemblies changed or other refueling modes. The first reactor cores of the new starting units may construct their first reactor cores directly based on the requirements of the needed refueling mode, so as to attain 18-month refueling mode, 12-month refueling mode with quarter of fuel assemblies changed, 12-month refueling mode with third of fuel assemblies changed or other refueling modes directly. For the fuel elements/assemblies joint configuration among the operating units, when there are several operating units which comprises the long-term and short-term operating units. The steps of the joint configuration of fuel elements/assemblies among the operating units are as follows: (B1) based on the equilibrium cycle configuration design, the spent fuel elements/assemblies having burned once which are original arranged in the peripheral place of the long-term operating units are replaced by the spent fuel elements/assemblies having burned twice, so as to obtain excess spent fuel elements/assemblies having burned once for reserve; (B2) the excess spent fuel elements/assemblies having burned once for reserve and a part of spent fuel elements/assemblies in the long-term operating units are arranged in the short-term operating units. Take the joint configuration design of nuclear fuel elements/assemblies of the operating units in 18-month refueling mode and 12-month refueling mode with quarter of fuel assemblies changed for example, further illustration is as follow. The operating units in 18-month refueling mode have the advantage of high available rate but the average discharge burnup of fuel is relatively low. The operating units in 12-month refueling mode with quarter of fuel assemblies changed have the advantage of high average discharge burnup of fuel. 12 spent fuel assemblies having burned once which are original arranged in the peripheral place of the operating units in 18-month refueling mode (as shown in FIG. 1) are replaced by the spent fuel assemblies having burned twice. The excess spent fuel assemblies having burned once for reverse and a part of the spent fuel assemblies that can not be used in the operating units refueling every 18 months but can be used in the operating units refueling every year with quarter of fuel assemblies changed are arranged into the reactor core of the operating units refueling every year with quarter of fuel assemblies changed for continuously combusting, so as to reduce the quantity of the new fuel assemblies required in the operating units refueling every year with quarter of fuel assemblies changed and to improve total fuel efficiency. So when the fuel elements/assemblies joint configuration is conducted by the reactor core of the operating units in long-term refueling mode and that in short-term refueling mode, the disadvantage of lower fuel efficiency of the operating units in long-term refueling mode can be partially compensated. The joint configuration design described above also can be applied to the units refueling when non-shutdown. Take the Pebble bed high temperature gas cooled reactor (HTR) units for example in the following to further illustrate the application of the present invention on the units refueling when non-shutdown. New fuel elements with relatively high enrichments same or similar to that in the equilibrium cycle are added additionally and gradually into the HTR units, and a large number of spent fuel elements with different burnups are discharged gradually for reverse; and then the spent fuel elements with different burnup having been discharged and in reserve and the new fuel elements with relatively high enrichments are mixed and loaded into the first core of the new starting units. The first cores of the new starting units may not use the fuel elements with a relatively low enrichment any longer wherein the discharge burnup is higher than that in the equilibrium cycle, so as to improve fuel efficiency significantly and save fuel costs. In the case where a number of fuel elements/assemblies appear damaged in one unit, a number of spent fuel elements/assemblies may be obtained from other operating units by the method for joint configuration design of nuclear power plant fuel elements/assemblies, as a result, the method for joint configuration design of nuclear power plant fuel elements/assemblies can also improve the ability of units to resist risk of fuel damage. The present invention is described by several specific embodiments. It should be understood for a person skilled in the art that various change and equivalent alternations can be made without departing from the scope of the present invention. Additionally, for specific circumstances or situation, various modifications can be made without departing from the scope of the present invention. So, the present invention is not limited to the specific embodiments in disclosure, and should comprise all the embodiments fallen into the scope of the appended claims of present invention. |
|
description | This application claims the benefit of U.S. Provisional Application No. 60/673,639, filed Apr. 21, 2005, the entire contents of which are incorporated herein by reference. This invention relates to surface cleaning of particulate contamination associated with microelectronic lithography and manufacturing. A particular technological challenge for implementing microelectronic lithography and manufacturing is the particulate contamination of surfaces associated with certain processes. For example, although pellicles are employed to protect the masks in conventional lithography, pellicles are not sufficiently transparent for wavelengths produced with extreme UV (EUV) lithography. Hence, particles on these masks print every exposure in this type of lithography. Moreover, various processes associated with EUV have a tendency to generate particles that contaminate other surfaces, in addition to masks. An example of an application to non-EUV lithography is nano-imprint lithography, which involves a mechanical process of pushing a UV-transmittable quartz mold into a thin film. When the mold is withdrawn, small amounts of contamination from the imprinted area may remain on the mold. When subsequent processing is performed with a contaminated mold, feature resolution decreases and there is potential for any film processed with the contaminated mold to be defective. Particle production can be limited to a certain extent, but it is unreasonable to expect this production to go to zero. As such, mitigation schemes and cleaning protocols of the various surfaces have been proposed. However, conventional wet cleaning with typical chemicals attacks the sensitive surfaces and, hence, degrades these surfaces overtime and therefore is not compatible with EUV or other low critical dimension semiconductor manufacturing. Recently, some research groups have proposed using laser-induced plasma cleaning systems to remove particulates from surfaces. In these systems, a laser scans across a surface and is focused near the particle to be removed such that breakdown of the gas occurs, resulting in a plasma. This plasma produces a shockwave that dislodges the particle. The shockwave, however, may damage the surface. Details of such systems are discussed in, for example, I. Varghese and C. Cetinkaya, J. Adhesion Sci. Technol., vol. 18, p. 295 (2004) and A. A. Busnaina, J. G. Park, J. M. Lee, and S. Y. Lou, IEEEISEMI Advanced Manufacturing Conference p. 41 (2003), the entire contents of which are incorporated herein by reference. Others have proposed using a plasma and gas flow based system to clean contaminated surfaces. This system generates reactive plasma that produces etching radicals, which reduces the adhesion force on the particles and the asperity height, which generally prevents rolling. A gas flow head is then moved across the surface to roll the particles off the surface. However, cleaning of patterned EUV masks is very difficult since rolling particles can fall into trenches or build up next to elevated features. Further details of such systems are described in Y. Momonoi, K. Yokogawa, M. Izawa, J. Vac. Sci. Technol. vol. B22, p. 268 (2004), the entire contents of which are incorporated herein by reference. Accordingly, in general, the present invention provides systems and methods for non-contact cleaning of particulate contamination of surfaces associated, for example, with microelectronic lithography and manufacturing. In one aspect, a system includes one or more sources that create a charge imbalance between a surface and particles that contaminate the surface with the possibility of a momentary electric field. This imbalance produces an electrostatic force that propels the particles off the surface. A power supply can be employed to produce a pulsed electrical bias to adjust the flux of electrons reaching the particles to control and optimize the rate at which the particles are charged. In certain embodiments, Malter-effect-like damage to the surface can be controlled through the use of plasma. An electron beam gun may be used to emit electrons at the surface and particles to create the charge imbalance. Alternatively, a plasma source may be employed to create plasma that produces the charge imbalance between the surface and particles. The plasma may be pulsed to adjust the flux of electrons reaching the particles to control and optimize the rate at which the particles are charged. Further features and advantages of the present invention will be apparent from the following description. Referring now to FIG. 1, a system embodying the principles of the present invention is illustrated therein and designated at 10. As its primary components, the system 10 includes an electron beam gun or emitter 12, a sample holder 14, and a plasma source 16, all contained in a chamber 18. The system 10 can be a stand alone system that cleans individual items, such as blank masks placed in the sample holder 14, or it can be integrated in a lithography system to clean patterned masks used, for example, in a stepper. Moreover, the system 10 can be integrated with the lithography system to clean the wafers during or between various steps of the lithography process. To use the system 10, an item with a contaminated surface is placed in a load lock 20 and then positioned in the holder 14 by an individual operator using a load lock manipulator 21 or by a robotic mechanism. The position of the contaminated surface in the chamber 18 can be manipulated through the use of adjusters located on the side of the chamber 18. Moreover, the holder 14 can be heated and rotated about 360°. After the item is in position, a gate valve 23 is closed to seal the chamber 18, and turbo pumps 25 are turned on to create a vacuum in the chamber 18. Pressure measurements in the chamber 18 may be performed through the use of a pressure probe 27. The electron beam gun 12 is then operated to expose particles on the surface with electrons to create a charge imbalance between the particles and the surface. This imbalance produces an electrostatic force that propels the particles off the surface. The system 10 may include a power supply and associated controller 30 connected to the isolated sample holder 14 to produce a pulsed electrical bias, such as a pulsed DC bias or RF bias, on the sample. The bias can be pulsed to adjust the flux of electrons reaching the particles and, hence, to control and optimize the rate at which the particles are charged. The electron beam gun 12 may have a current density of between about 1 and 500 mA/cm2 depending on the size of the particle being removed and the spot size of the beam desired. A current density of about 25 mA/cm2 and a spot size of about 2 cm2 are preferable. The system 10 may include a Faraday cup 32 and associated manipulator to measure the electron beam current. The electron beam gun 12 can produce a wide area beam to flood the surface with electrons, which affords a relatively quick and efficient method to effectively clean the entire surface. Alternatively, the electron beam gun 12 can produce a collimated beam that is steered to target specific areas of the surface. During the exposure of the particles and surface, the electron beam gun 12 can be turned on continuously, or, in some applications, intermittently. The cycle time the electron beam gun is turned on may be varied between about a few seconds to a couple of minutes. Thus, the electro beam gun 12 can be modulated in amplitude or time to produce staged levels of charge imbalance on the particles. In addition, the pulse length of the electrical bias produced by the power supply and associated controller 30 can be varied from micro-seconds to milliseconds. A particular feature of the system 10 is that low currents and short pulse times can be used to first remove a particular size particle, and then the current or DC bias duration and pulse length can be varied to remove other size particles. Further, the overall time can be varied to remove each size of particle or composition class of materials in turn. Gas flow can also be present and can be varied in order to complement this effect and prevent particles from reattaching to the surface being cleaned. In certain implementations, the plasma source 16 produces plasma based on an inert gas, such as Ar or He, or H to mitigate sputtering damage to the surface. For example, the plasma may contact the particles and surface to reduce Malter-effect-like damage to the surface. The plasma source 16 rather than the electron beam gun may be used to produce plasma 34 to create the charge imbalance. The plasma source 16 may be operated continuously. Alternatively, the plasma source 16 may be pulsed. That is, the power supply for the plasma source 16 may be turned off and on, for example, under the direction of a controller, to control and optimize the rate at which the charge imbalance is created. The pulse length of the pulsed plasma can be varied from micro-seconds to milliseconds. A set of magnetic coils 36 may be employed to guide plasma 34 produced by the plasma source 16. A Langmuir probe 38 may be employed to determine properties of the plasma 34, such as its temperature and density. In various implementations, the pulsed power supply and controller 30 or the pulsed power supply of the plasma source 16 can operate at a few hundred to about 200 kHz and can alter the pulsed waveform to get different charge rates and total charge delivered to the target. A rate of 50 kHz for about 3 minutes using a 50% duty cycle square wave pulse is preferable. The system 10 may also include a gas supply and an associated controller to provide the flow of gas, such as Ar, across the surface to enhance the cleaning capabilities of the system 10, and a gas analyzer 40, such as a residual gas analyzer (RGA), to characterize the gasses in the chamber 18. Reactive gasses can also be used such as chlorine. A particular feature of the system 10 is its ability to remove particulates on the submicron, in particular, the nanometer, scale that come to reside on the surfaces associated, for example, with EUV. To help understand this feature, the Van der Waals force, F, that holds a particle to the surface of the optic material is defined as:F=0.75*π*WA*D (1)where D is the diameter of the particle and WA is given by: W A = A 12 * π * z 0 2 ( 2 ) where A is the Hamaker's constant and z0 is the distance between the particle and the surface. Using polystyrene as an example, A is 12.2×10−20 J and z0 is approximately 4 Angstroms. Thus, for polystyrene particles with a diameter of about 200 nm, F≈10 nN. The electrostatic force on the particle can be calculated from: F = q p 2 4 πɛ 0 R 2 ( 3 ) Hence, by knowing the approximate particle size, and the previously calculated resultant Van der Waals force, the amount of electrostatic charge qp, can be deduced. For a polystyrene particle with a diameter of about 200 nm, only 1600 elementary charges are needed for F=10 nN. Also, as R is reduced, qp is reduced proportionally, and there is, therefore, no theoretical minimum limit on the size of the particle removed. When the system 10 is in use, charge is deposited onto contaminant particles as well as the substrate upon which the particles reside either by the broad beam electron gun or emitter 12 or through the use of plasma. When charge is first deposited on a non-conducting particle, an equal and opposite image charge is produced under the surface as shown in FIG. 2. When charge is deposited by use of plasma, the ion flux to the surface of the substrate equals the ion flux attributable to electron repulsion in the plasma sheath as shown in FIG. 3. Most particles have sufficient conductivity to the substrate, and a charge balance is reached. Adding a positive pulsed on non-pulsed bias to the surface has three effects: (1) it adds additional electrons to the particle, (2) it produces negative trapped charge, for example, in the SiO2 native oxide on such a substrate surface thereby screening the original image charge, and (3) the moving charges themselves may screen the image charge. All three effects add to the electrostatic repulsion between the particle and substrate. FIG. 4 shows these effects with a plasma source used as the electron source. Note that the electron beam gun or emitter 12 produces the same response. This electrostatic repulsion can be further increased by pulsing the plasma. When the power input to the plasma is turned off, electrons quickly exit the plasma leaving the heavier ions behind. Therefore, the negative charge on the particle is momentarily increased again while an electric field is present. The plasma “afterglow” after the electrons have left the plasma and before the ions have recombined or hit the surface has the effect of being an electrostatic pump, as illustrated in FIG. 5. The system 10 can be used to clean many types of surfaces. For example, in some applications, as described previously, the surfaces of mask blanks or patterned masks used in lithography processes may be cleaned with the system 10. The surfaces may be those of semi-conductor wafers. The mask or wafers may be charged in a vacuum chamber during the lithography process, or they may be charged during the process after removal from the vacuum chamber by, for example, a robot. The system 10 may be a standalone system, or one or more systems 10 may be integrated with a lithography system. The surfaces cleaned may be related to biochips. The surfaces may be associated with conductors, semi-conductors, or non-conductors. The system 10 may also be used in conjunction with imprint lithography. Other applications include cleaning MEMS and NEMS devices or any micro-tech or nano-tech structure or surface where particle removal and surface cleanliness is important. I. Experiment Setup The system 10 was setup for non-contact cleaning of surfaces associated with EUV. Particle application was accomplished through the use of a diluted solution of ˜220 nm diameter polystyrene nanoparticles suspended in isopropyl or methyl alcohol. The use of an alcohol minimizes the polarity of the solution, which can cause agglutination of particles on the surface of the sample during application. The alcohol solution is put into the reservoir of a commercial nebulizer. The nebulizer is used to transform a liquid solution into an aerosol solution by forcing compressed air into the reservoir. The force of the air is such that it breaks the liquid into tiny mist-like particles that consist of the alcohol and polystyrene particles in suspension. The entire nebulizer reservoir is immersed in an ultrasonic bath. Because of the size of the polystyrene nanoparticles, Van der Waals forces dominate and result in the particles self-agglutinating into clumps, which is undesirable since there is little clumping that occurs in EUV processes. The ultrasonic bath provides a mechanism to continuously overcome the Van der Waals forces and, therefore, to keep the particles in suspension. This provides a more uniform application of the solution on to the sample material and results in very little agglutination of the nanoparticles into clumps on the surface of the sample. The mist of the nebulizer is directed onto the sample at a specific distance and time that has been calibrated to yield a specific and repeatable particle density. To quantify the non-contact cleaning of the samples, SEM analysis of the various samples were performed in conjunction with the use of a commercial statistical analysis program, Image J, by W. S. Rasband, at the National Institutes of Health in Bethesda, Ma., to yield an average number of particles per square micron. This begins by starting with a standard SEM picture and removing the grey background. A region is then selected and the non-black and black pixels are counted. In the SEM picture, the black or darker pixels correspond to something that is lower in height or level than a lighter pixel. Thus, particles on the surface show up as white dots, which is the basis behind determining how to count particles and pixels. This procedure was performed several times on the same sample at various locations to provide a statistical average. II. Experiment The samples used for this investigation were bare (1,1,1) Si wafer samples with an area of about 1 cm2. Using (1,1,1) Si wafers are a viable test standard to investigate non-contact cleaning of contaminants and to develop optimal protocols before running experiments with other more expensive materials. Si samples were nebulized with ˜220 nm particles to provide a coverage of 0.46±0.25 particles per square micron, as shown in FIG. 6. This was the control measurement to which subsequent “cleaned” samples were compared. While holding the sample at ground potential in vacuum, a broad base electron beam at 200 eV with a current of 65 mA was applied for 60 secs, producing the results (0.092±0.074 particles per square micron) shown in FIG. 7. Another sample was also put in vacuum and a pulsed DC bias of ˜200 V at 50 kHz and a broad electron beam with the same characteristics were applied, producing the results (0.50±0.21 particles per square micron) shown in FIG. 8. III. Results In comparison to the original samples (FIG. 6), the application of a broad electron beam resulted in approximately 80% reduction in particle density (FIG. 7). To show that the removal of the particles was a function of charging the particles, in the subsequent experiment (FIG. 8), a pulsed bias was provided at the same time to eliminate the effective charging on the particles. As a result, there was effectively no reduction in particle density, which illustrates the successful non-contact removal effect that can be achieved through electrostatic charging of the polystyrene. Thus, it was shown that about 80% of these ˜220 nm particles were removed by electrostatic repulsion for this configuration. I. Experiment Setup The particle application to the samples for this experiment were the same as that for Example A, except that equal numbers of ˜30 nm, ˜80 nm, and ˜220 nm particles were used in solution instead of entirely ˜220 nm particles. II. Experiment For this experiment, a sample was exposed to a 2,000 W Ar plasma for ten minutes with a 10 mA constant bias on the sample. At the end of 10 minutes, the sample was flipped over before turning the plasma off to prevent re-adhesion of particles to the surface. III. Results The visual results are compared in FIG. 9. The test sample (right image) compared to a control sample (left image) exhibits faint residue remaining at the edges of the clusters of particles removed. I. Experimental Setup The particle application to the samples for this experiment were the same as that for Example B. II. Experiment For this experiment, a sample was exposed to a 2,000 W Ar plasma for ten minutes with 10 mA constant bias on the sample. The plasma was pulsed once a minute for 10 minutes, then the sample was flipped over before turning the plasma off again to lessen the likeliness of the re-adhesion of particles. III. Results The visual results of the control sample (left image) and the test sample (right image) are shown in FIG. 10. FIG. 11 illustrates the results at a higher magnification showing the removal of ˜30 nm and ˜80 nm particles. The test samples in FIGS. 10 and 11 show no particles present but slight chemical residue from the particle application process. The present invention has been described with reference to specific embodiments, which are provided only for exemplification and are not to be construed as limiting the scope of the invention as defined by the following claims. |
|
summary | ||
abstract | A collimator for X-ray imaging apparatus is provided. The collimator includes a collimator housing including a tube flange, a tube, a locating ring configured to be mounted at an outlet of the tube flange, and at least one tongue set fixed on the locating ring. An outstretching direction of the tongue is towards a center of the locating ring. |
|
052873951 | abstract | A double crystal monochromator including two identical, parallel crystals, each of which is cut such that the normal to the diffraction planes of interest makes an angle less than 90 degrees with the surface normal. Diffraction is symmetric, regardless of whether the crystals are symmetrically or asymmetrically cut, enabling operation of the monochromator with a fixed plane of diffraction. As a result of the inclination of the crystal surface, an incident beam has a footprint area which is elongated both vertically and horizontally when compared to that of the conventional monochromator, reducing the heat flux of the incident beam and enabling more efficient surface cooling. Because after inclination of the crystal only a fraction of thermal distortion lies in the diffraction plane, slope errors and the resultant misorientation of the diffracted beam are reduced. |
051005859 | claims | 1. A process of recovering strontium and/or technetium values from an aqueous nitric acid feed solution containing nitric acid and fission product values comprising: contacting the aqueous solution containing of about 1-6 molar nitric acid and technetium and strontium values with an extractant consisting of a crown ether in a diluent, the diluent being an organic compound which is insoluble in water, is capable of dissolving at least 0.5M molar water, and in which the ether is soluble, maintaining the contact for a period of time sufficient for the strontium and technetium values in the feed solution to be taken up by the extractant, and separating the extractant from the feed solution, thereby separating the strontium and technetium values from the aqueous nitric acid solution. contacting the nitric acid solution of about 1 to 6 molar nitric acid with an extractant consisting of from 0.1 to 0.5M of a crown ether in a diluent, the crown ether being selected from the group consisting of dicyclohexano-18-crown-6, bis-methylcyclohexano-18-crown-6, and 4,4'(5')[(t-butyl)cyclohexano]-18-crown-6, and the diluent being selected from the group consisting of 1-octanol, 1-heptanol, 1-decanol, octanoic acid, heptanoic acid, hexanoic acid, 2-hexanone, 4-methyl-2-pentanone, butyl acetate and amyl acetate, maintaining this contact for a period of time sufficient for the strontium and technetium values in the feed solution to be taken up by the extractant, and separating the extractant from the feed solution, thereby separating the strontium and technetium values from the nitric acid feed solution. 2. The process of claim 1 wherein the feed solution contains up to 6 molar nitric acid. 3. The process of claim 2 wherein the extractant is from about 0.1 to 0.5M in crown ether. 4. The process of claim 3 wherein the diluent is a member selected from the group consisting of alcohols, ketones, carboxylic acids and esters. 5. The process of claim 4 wherein the crown ether has the formula: 4,4'(5')[(R,R')cyclohexano]-18-crown-6, where R and R' are one or more members selected from the group consisting of H, methyl, t-butyl, and heptyl. 6. The process of claim 4 wherein the diluent is an alcohol selected from the group consisting of 1-octanol, 1-heptanol and 1-decanol. 7. The process of claim 4 wherein the diluent is a carboxylic acid selected from the group consisting of octanoic acid, heptanoic acid and hexanoic acid. 8. The process of claim 4 wherein the diluent is a ketone selected from the group consisting of 2-hexanone and 4-methyl-2-pentone. 9. The process of claim 4 wherein the diluent is a ester selected from the group consisting of butyl acetate and amyl acetate. 10. The process of claim 6 wherein the crown ether is selected from the group consisting of dicyclohexano-18-crown-6, bis-methylcyclohexano-18-crown-6, and bis 4,4'(5')[(t-butyl)cyclohexano]-18-crown-6. 11. The process of claim 7 wherein the crown ether is selected from the group consisting of dicyclohexano-18-crown-6, bis-methylcyclohexano-18-crown-6, and bis 4,4'(5')[(t-butyl)cyclohexano]-18-crown-6. 12. The process of claim 8 wherein the crown ether is selected from the group consisting of dicyclohexano-18-crown-6, bis-methylcyclohexano-18-crown-6, and bis 4,4'(5')[(t-butyl)cyclohexano]-18-crown-6. 13. The process of claim 9 wherein the crown ether is selected from the group consisting of dicyclohexano-18-crown-6, bis-methylcyclohexano-18-crown-6, and bis 4,4'(5')[(t-butyl)cyclohexano]-18-crown-6. 14. The process of claim 6 wherein the diluent is 1-octanol and the crown ether is selected from the group consisting of dicyclohexano-18-crown-6, bis-methylcyclohexano-18-crown-6, and bis 4,4'(5')[(t-butyl)cyclohexano]-18-crown-6. 15. The process of claim 14 wherein the crown ether is bis 4,4'(5')[(t-butyl)cyclohexano]-18-crown-6. 16. A process for recovering strontium and/or technetium values from an aqueous feed solution containing nitric acid together with other fission product values comprising: |
039376532 | abstract | A diagrid for supporting a reactor core having a vertical axis is constituted by a box structure consisting of a flat and horizontal upper plate and a domical lower plate. The upper plate has a circular flange parallel to a peripheral rim of the lower domical plate, the rim and the flange being braced relative to each other. The flange of the upper plate rests on a ring-girder which surrounds the box structure and provides a peripheral side restraint whilst the flat rim of the lower domical plate rests on a flat circular flange of a support which forms a conical downward extension of the domical plate and is rigidly fixed to the reactor vessel. |
058898328 | abstract | The control cluster has a spider constituted by a hub connectable to a drive shaft. Fins radiate from a bottom portion of the hub and provided with vertical fingers distributed in a regular array. It also includes rods provided with plugs that are releasably fixed to the fingers. The fins, their fingers, and at least the bottom portion of the hub constituting a single-piece part obtained by molding or by electro-machining. |
claims | 1. An illumination device comprising a radiation sourcehaving at least one light-emitting diode;a control unit; anda radiation receiving unit;wherein the radiation receiving unit is provided, during operation of the illumination device for receiving both a radiation emitted by the radiation source and a reference radiation and for generating a measurement signal upon receiving the radiation from the radiation source and a reference signal upon receiving the reference radiation, andwherein an operating point for the radiation source is tunable by means of the control unit in a manner concurrently dependent on the measurement signal and the reference signal,wherein the measurement of the reference signal and the measurement signal take place at the same time. 2. The illumination device as claimed in claim 1, wherein a color locus of the radiation from the radiation source is tunable by means of the control unit. 3. The illumination device as claimed in claim 2, wherein the color locus of the radiation from the radiation source is different from a color locus of the reference radiation in a targeted manner. 4. The illumination device as claimed in claim 2, wherein the color locus of the radiation from the radiation source corresponds to a color locus of the reference radiation. 5. The illumination device as claimed in claim 1, which comprises an electrically operable reference radiation source provided for generating the reference radiation. 6. The illumination device as claimed in claim 5, wherein the reference radiation source is nominally provided for operation with a rated power and the reference radiation source is operable below the rated power. 7. The illumination device as claimed in claim 1, which has a radiation exit surface. 8. The illumination device as claimed in claim 7, wherein the reference radiation source is arranged outside an optical beam path from the radiation source to the radiation exit surface. 9. The illumination device as claimed in claim 8, wherein the reference radiation source comprises an incandescent lamp or a gas discharge lamp. 10. The illumination device as claimed in claim 8, wherein the reference radiation source comprises an optoelectronic semiconductor component. 11. The illumination device as claimed in claim 1, wherein the illumination device comprises a phosphorescent material provided for generating the reference radiation. 12. The illumination device as claimed in claim 11, wherein the phosphorescent material is excited during operation of the illumination device and the reference radiation is generated by means of the persistence of the phosphorescent material. 13. The illumination device as claimed in claim 11, wherein the phosphorescent material is arranged and formed in such a way that it can be optically excited by means of the radiation source. 14. The illumination device as claimed in claim 11, wherein the phosphorescent material is arranged and formed in such a way that it can be excited independently of the radiation source. 15. The illumination device as claimed in claim 5, wherein the illumination device comprises a phosphorescent material which is provided for generating the reference radiation and which can be optically excited by means of the reference radiation source. 16. The illumination device as claimed in claim 11, wherein the phosphorescent material is integrated into the radiation source. 17. The illumination device as claimed in claim 11, wherein the phosphorescent material is formed separately from the radiation source. 18. The illumination device as claimed in claim 7, wherein radiation emitted by the radiation source is directed through the radiation exit surface at least partly by means of a reflective surface. 19. The illumination device as claimed in claim 18, wherein the reflective surface is embodied such that it is partly transparent to radiation emitted by the radiation source. 20. The illumination device as claimed in claim 18, wherein the radiation receiving unit is arranged on that side of the reflective surface which is remote from the radiation exit surface. 21. The illumination device as claimed in claim 1, wherein the radiation receiving unit comprises a radiation receiver which is sensitive over the visible spectral range. 22. The illumination device as claimed in claim 1, wherein the radiation receiving unit comprises a respective radiation receiver for at least two mutually different spectral ranges. 23. The illumination device as claimed in claim 22, wherein the radiation receiving unit is formed by means of radiation receivers formed in discrete fashion. 24. The illumination device as claimed in claim 22, wherein the radiation receiving unit is formed by means of radiation receivers formed in monolithically integrated fashion. 25. A method for adapting an emission characteristic of an illumination device to a predetermined emission characteristic, comprising the following steps:a) receiving a radiation from a radiation source of the illumination device by means of a radiation receiving unit and generating a measurement signal;b) feeding the measurement signal to a control unit;c) receiving a reference radiation by means of the radiation receiving unit and generating a reference signal;d) feeding the reference signal to the control unit; ande) setting an operating point for the radiation source by means of the control unit in a manner concurrently dependent on the measurement signal and the reference signal, wherein the measurement of the reference signal and the measurement signal take place at the same time. 26. The method as claimed in claim 25, wherein a sensitivity of the radiation receiving unit is calibrated by means of the reference radiation. 27. The method as claimed in claim 25, wherein a change in the sensitivity of the radiation receiving unit on account of at least one of aging of the radiation receiving unit or a change in temperature of the radiation receiving unit is monitored by means of the reference radiation. 28. The method as claimed in claim 25, wherein the color locus of the radiation generated by the radiation source is determined by means of the radiation receiving unit. 29. The method as claimed in claim 25, wherein the radiation source comprises at least two radiation emitters which are driven separately by the control unit. 30. The method as claimed in claim 29, wherein the radiation emitters emit radiation in mutually different spectral ranges. 31. The method as claimed in claim 29, wherein the radiation emitters are operated simultaneously for determining the color locus of the radiation source. 32. The method as claimed in claim 29, wherein the radiation emitters are operated successively for determining the color locus of the radiation source. 33. The method as claimed in claim 25, wherein the operating point for the radiation source is determined from the measurement signal and the reference signal by means of an arithmetic operation. 34. The method as claimed in claim 25, wherein the operating point is set by means of the control unit in such a way that a change in the emission characteristic that is induced by a change in temperature of the radiation source is at least partly compensated for. 35. The method as claimed in claim 25, wherein the operating point is set by means of the control unit in such a way that a change in the emission characteristic that is induced by aging of the radiation source is at least partly compensated for. 36. The method as claimed in claim 25, wherein the reference radiation is generated by means of a phosphorescent material. 37. The method as claimed in claim 36, wherein the phosphorescent material is excited and the reference radiation is generated by means of a persistence of the phosphorescent material. 38. The method as claimed in claim 36, wherein the phosphorescent material is optically excited. 39. The method as claimed in claim 36, wherein the phosphorescent material is excited by the radiation source. 40. The method as claimed in claim 29, wherein the reference signal is generated while the radiation source is switched off. 41. The method as claimed in claim 25, wherein the reference radiation is generated in such a way that an aging-dictated alteration of the emission properties of the reference radiation is smaller than aging-dictated alteration of the emission properties of the radiation source. 42. The method as claimed in claim 25, wherein the illumination device comprises a reference radiation source which is operated electrically. 43. The method as claimed in claim 42, wherein the reference radiation source is nominally provided for operation with a rated power and the reference radiation source is operated below the rated power. 44. The method as claimed in claim 42, wherein the reference radiation source is operated for a shorter time than the radiation source during operation of the illumination device. 45. The method as claimed in claim 36, wherein the illumination device comprises a reference radiation source which is operated electrically and wherein the phosphorescent material is excited by the reference radiation source. 46. The method as claimed in claim 25, wherein the measurement signal and/or the reference signal are stored in the control unit. 47. The method as claimed in claim 25, wherein the illumination devicecomprises a radiation source having at least one light-emitting diode;a control unit; anda radiation receiving unit;wherein the radiation receiving unit is provided, during operation of the illumination device for receiving both a radiation emitted by the radiation source and a reference radiation and for generating a measurement signal upon receiving the radiation from the radiation source and a reference signal upon receiving the reference radiation, andwherein an operating point for the radiation source is tunable by means of the control unit in a manner concurrently dependent on the measurement signal and the reference signal. |
|
description | The present invention relates to a technique for substantially decreasing generation of a sodium nitrate liquid waste, that is a secondary waste, by reductively decomposing a sodium nitrate liquid waste produced as a secondary waste through a wet reprocessing process of a spent nuclear fuel and recycling the obtained sodium salt(s) selected from sodium hydroxide, sodium hydrogencarbonate, and sodium carbonate. Purex process that has been conventionally used as a reprocessing technique for a light-water reactor fuel is known as a wet reprocessing process of spent nuclear fuels. In this process, after shearing a spent nuclear fuel, small pieces of the fuel are dissolved in nitric acid [dissolution step]. The nitric acid solution includes uranium, plutonium and FP (fission products), and insoluble residues of FP and solid impurities such as cladding tube chips produced during shearing. Thus, these residues and impurities are removed, and the nitric acid concentration and the like are adjusted [clarification and adjustment step]. This nitric acid solution is brought into contact with a mixed organic solvent of n-dodecane and TBP (tributyl phosphate) to extract uranium and plutonium into the organic solvent phase [extraction step], and to subject FP left in the nitric acid aqueous phase to vitrification as a high level liquid waste. The organic solvent phase containing uranium and plutonium is brought into contact with a nitric acid aqueous phase containing uranium (IV) or HAN (hydroxylamine nitrate) to back extract plutonium into the nitric acid aqueous phase and leave uranium in the organic solvent phase [distribution step]. The organic solvent phase containing uranium is further brought into contact with a diluted nitric acid solution to back extract uranium into the nitric acid aqueous phase. The obtained nitric acid solution containing uranium and the obtained nitric acid solution containing plutonium are subjected respectively to extraction, washing, back extraction, and condensation to remove impurities such as FP [purification step]. The purified nitric acid solution containing uranium and the purified nitric acid solution containing plutonium are also denitrated to recover a uranium oxide and a plutonium oxide respectively [denitration step]. In addition, development of reprocessing techniques represented by RETF (Recycle Equipment Test Facility), advanced wet reprocessing processes, and advanced wet reprocessing processes adopting direct extraction have been proposed as improved Purex processes (Non-Patent Documents 1 and 2). The development of reprocessing techniques for fast reactor fuels in RETF has started on the basis of Purex process which has been put to practical use as a reprocessing technique of a spent nuclear fuel from a light-water reactor. Since the performance of each step of dissolution, clarification, and extraction needs to be improved in response to increasing FP contents due to increased fast reactor fuel burnup, the equipments such as continuous dissolvers, centrifugal clarifiers, and centrifugal extractors have been newly developed. However, separation of uranium and plutonium from a nitric acid solution of a spent nuclear fuel is carried out in accordance with Purex process. In the advanced wet reprocessing process, some of a large amount of uranium included in a nitric acid solution of a spent nuclear fuel is separated and recovered in advance by an uranium crystallization method in which the temperature dependence of the uranium solubility is used to precipitate and separate uranium, thereby reducing the amount of nuclear materials to be treated in a subsequent solvent extraction step and the following steps. In order to use as a MOX (mixed oxide) fuel, an uranium oxide and a plutonium oxide that are recovered in conventional Purex process, have been mixed at an appropriate ratio. However, the uranium/plutonium ratio in a nitric acid solution of a spent nuclear fuel to be subjected to Purex process is controlled in advance to an appropriate ratio suitable for a MOX fuel, and uranium and plutonium are extracted into an organic solvent and then back extracted all together from the organic solvent, thereby allowing only an uranium/plutonium mixed oxide of the appropriate ratio to be obtained. Such a process has been proposed in the advanced wet reprocessing process. The advanced wet reprocessing processes adopting direct extraction has been proposed as one of alternative techniques to the advanced wet reprocessing process, where a solvent in which nitric acid and TBP form a complex (TBP nitric acid complex) is directly brought into contact with a solid spent nuclear fuel, thereby selectively recovering uranium and plutonium. This process can be simplified as compared with the advanced wet reprocessing process, and reduces nitric acid for dissolution and the amount of liquid waste generated from an extraction step, and thus the reduction of the step for concentrating a high level liquid waste can be thus expected. The above-described wet reprocessing processes include a dissolution step for dissolving a spent nuclear fuel in nitric acid, and nitric acid in washing liquids produced in separation or distribution of and a purification process of uranium or plutonium is mostly recovered by a nitric acid recovery system such as concentration by evaporation and reused, while some of the nitric acid results in excess nitric acid. This excess nitric acid is neutralized with a sodium salt selected from sodium hydroxide, sodium hydrogencarbonate and sodium carbonate, thereby producing a sodium nitrate liquid waste. It is to be noted that the term “sodium salt” as used herein refers to sodium hydroxide, sodium hydrogencarbonate or sodium carbonate. It is often the case that liquid wastes produced in the analyses carried out in each step of the wet reprocessing process are also nitrate forms although occurring in slight amounts, and if these liquid wastes are acidic wastes, the wastes are subjected to neutralization processing with a sodium salt, thereby producing a sodium nitrate liquid waste. Further, an off-gas produced in dissolving the spent nuclear fuel in the nitric acid includes nitrogen oxides, a slight amount of radioactive components and the like. The off-gas is thus washed with an alkali solution of a sodium salt (alkali scrubbing) in order to remove the nitrogen oxides, the radioactive components and the like, thereby producing a sodium nitrate liquid waste as the washing liquid waste. In addition to this, an organic solvent degraded due to radiation is washed with a sodium salt in order to remove the degraded component. The liquid waste of the sodium salt is neutralized with nitric acid to produce a sodium nitrate liquid waste. These sodium nitrate liquid wastes produced as secondary wastes through the wet reprocessing process are subjected to the processing such as concentration by evaporation, and the condensate liquid is discharged. Since the concentrated liquid is a low-level radioactive waste, the radioactive waste is subjected to vitrification, cement solidification or bituminization, or kept liquid, powdered, or pelletized for intermediate storage. On the other hand, various reduction methods have been conventionally proposed, which may be catalytic reduction methods in which nitrate-nitrogen in liquid waste water containing nitrate-nitrogen is reduced to nitrogen gas with a reducing agent and catalyst or supercritical reduction methods in which the nitrate-nitrogen is reduced to nitrogen gas with a reducing agent under supercritical conditions where water serves as a supercritical fluid. Methods using hydrogen as the reducing agent and methods using a reducing agent other than hydrogen are known as the catalytic reduction methods. The methods using hydrogen as the reducing agent include, for example, a method in which nitrate-nitrogen is reduced to nitrogen with a zeolite catalyst in the presence of hydrogen (Non-Patent Document 3). The methods using a reducing agent other than hydrogen include, for example, a method in which a reducing agent such as hydrazine is added to waste water containing nitrate-nitrogen, and the waste water is brought into contact with a sponge copper catalyst to reduce the nitrate-nitrogen to nitrite-nitrogen, and then further brought into contact with a palladium catalyst with hydrazine or the like to reduce the nitrite-nitrogen to nitrogen gas (Patent Document 1). The supercritical reduction methods include, for example, a method in which a nitrogen in a nitrate is reduced to nitrogen gas with a reducing agent such as an alcohol, ammonia, carbohydrate, formic acid, or oxalic acid under conditions where water serves as a supercritical fluid, that is, at a temperature and a pressure equal to or more than the critical point (374° C., 22 MPa) of water (Patent Document 2). Non-Patent Document 1: “Development on FBR Fuel Reprocessing Technology”, Japan Society of Mechanical Engineers (No. 96-3), 5th Power and Energy Technology Symposium '96, Collected Papers for Lectures Non-Patent Document 2: “Component Technologies Development of Reprocessing System—Advanced Aqueous Reprocessing Process Technologies Development—”, Report from Japan Nuclear Cycle Development Institute, No. 24 Separate Volume, 153-164, November 2004 Non-Patent Document 3: “Development of Wet Reductive Decomposition Processing Technique for Nitrate-Nitrogen in Wastewater”, Report on Completion of Program for Promotion of Technology Development from 2002 to 2004, I-26, RITE-Wakamatsu Second Laboratory, Research Institute of Innovative Technology for the Earth, March 2005 Patent Document 1: Japanese Patent Laid-Open No. 2003-126872 Patent Document 2: Japanese Patent Laid-Open No. 2005-241531 As described above, the sodium nitrate liquid wastes produced as secondary wastes through the wet reprocessing process of a spent nuclear fuel are subjected to processing such as concentration by evaporation, and the condensate liquid is discharged. Since the concentrated liquid is a low-level radioactive waste, the radioactive waste is subjected to vitrification, cement solidification or bituminization for storage, or kept liquid, powdered, or pelletized for intermediate storage. However, when a spent nuclear fuel is subjected to wet reprocessing, sodium nitrate liquid waste that are secondary waste are inevitably produced, and the amount of radioactive waste to be processed correspondingly increases. This issue becomes a big problem when a wet reprocessing process is carried out. It is thus an object of the present invention to provide a novel and improved method that is capable of substantially reducing the amount of radioactive waste generated due to a sodium nitrate liquid waste by reductively decomposing a sodium nitrate liquid waste produced through a wet reprocessing process of a spent nuclear fuel and recovering and reusing a sodium salt as a decomposition product. The inventors have found that generation of a secondary waste due to a sodium nitrate can be effectively reduced by applying a conventionally proposed catalytic reduction method or a supercritical reduction method to processing of a sodium nitrate liquid waste produced through a wet reprocessing process of a spent nuclear fuel or of a sodium nitrate liquid waste produced through washing of an organic solvent used in a wet reprocessing process to reductively decompose sodium nitrate, and completed the present invention. More specifically, an embodiment of the present invention is a sodium salt recycling process for use in wet reprocessing process of a spent nuclear fuel, characterized by comprising: a neutralization step in which a nitric acid liquid waste or an off-gas having nitric acid dissolved therein which is produced through a wet reprocessing process including a dissolution step for dissolving a spent nuclear fuel in nitric acid is neutralized by adding or contacting one or more sodium salts selected from sodium hydroxide, sodium hydrogencarbonate and sodium carbonate to or with the nitric acid liquid waste or the off-gas, thereby yielding a sodium nitrate liquid waste; a sodium nitrate-decomposition step in which the sodium nitrate liquid waste obtained in the neutralization step is reductively decomposed with a reducing agent, thereby decomposing sodium nitrate into a nitrogen gas and the sodium salt(s); and a recycle step for recycling the sodium salt(s) produced in the sodium nitrate-decomposition step into the neutralization step or the wet reprocessing process. If necessary, an evaporative concentration step in which the sodium nitrate liquid waste obtained in the neutralization step is concentrated by evaporation may be provided, and the concentrated sodium nitrate liquid waste obtained in the evaporative concentration step may be reductively decomposed in the sodium nitrate-decomposition step. Another embodiment of the present invention is a sodium salt recycling process for use in wet reprocessing process of a spent nuclear fuel, characterized by comprising: an organic solvent washing step in which an organic solvent used in a wet reprocessing process of a spent nuclear fuel is washed with one or more sodium salts selected from sodium hydroxide, sodium hydrogencarbonate and sodium carbonate; a neutralization step in which a liquid waste of the sodium salt(s) produced in the organic solvent washing step is neutralized with nitric acid, thereby yielding a sodium nitrate liquid waste; a sodium nitrate-decomposition step in which the sodium nitrate liquid waste produced in the neutralization step is reductively decomposed with a reducing agent, thereby decomposing sodium nitrate into a nitrogen gas and the sodium salt(s); and a recycle step for recycling the sodium salt(s) produced in the sodium nitrate-decomposition step into the organic solvent washing step or the wet reprocessing process. If necessary, an evaporative concentration step in which the sodium nitrate liquid waste obtained in the neutralization step is concentrated by evaporation may be provided, and the concentrated sodium nitrate liquid waste obtained in the evaporative concentration step may be reductively decomposed in the sodium nitrate-decomposition step. The reductive decomposition in the sodium nitrate-decomposition step can preferably employ reductive decomposition that is carried out with both a reducing agent and a catalyst, or reductive decomposition that is carried out with a reducing agent under supercritical conditions where water serves as a supercritical fluid. Further, it is preferable that a part of the sodium salt(s) to be recycled in the recycle step can be sent and mixed in a solidified substance of a radioactive waste for solidification or in a vitrified substance of a high-level radioactive waste for vitrification. According to the present invention, the reductive decomposition of the sodium nitrate liquid waste produced as a secondary waste through a wet reprocessing process of a spent nuclear fuel allows nitrate ions in the liquid waste to be reduced into a nitrogen gas and released into the air, and allows sodium ions to be recovered and recycled in a reusable form as sodium hydroxide, sodium hydrogencarbonate, or sodium carbonate, thereby allowing the amount of waste generated due to the sodium nitrate liquid waste to be substantially decreased. Further, the amount of liquid waste to be decomposed can be decreased by concentrating by evaporation the sodium nitrate liquid waste obtained in the neutralization step to give a concentrated sodium nitrate liquid waste, and then reductively decomposing the concentrated sodium nitrate liquid waste, thus allowing the capacity of processing equipment to be decreased and allowing the processing time to be decreased. Furthermore, a part of the sodium salt recycled in the recycle step is brought and mixed in a solidified substance of a radioactive waste for solidification, thereby allowing accumulation of radioactivity in the recycled sodium salt(s) to be reduced. In particular, when some of the sodium salt(s) to be recycled is mixed in a vitrified waste, the sodium salt(s) can be used as Na2O in borosilicate glass components, thus allowing the corresponding amount of new sodium salt(s) to be used in the neutralization step and allowing the amount of waste generated due to the sodium nitrate liquid waste to be cleared. FIG. 1 is a process chart illustrating an example according to an embodiment of the present invention. It is to be noted that although FIG. 1 and the following description illustrate only sodium hydroxide as a typical example of a sodium salt produced by reductive decomposition of nitric acid and of a sodium salt to be recycled, the present invention includes cases in which the sodium salt is sodium hydrogencarbonate or sodium carbonate, or is a mixture of one or more of sodium hydroxide, sodium hydrogencarbonate and sodium carbonate. The wet reprocessing process of a spent nuclear fuel is not limited to Purex process as long as the wet reprocessing process is a reprocessing process including a dissolution step for dissolving a spent nuclear fuel in nitric acid, reprocessing techniques represented by RETF, advanced wet reprocessing processes, and advanced wet processes adopting direct extraction, which have been introduced as prior art, and the like can be also employed. Nitric acid generated in a separation or distribution process of uranium or plutonium and nitric acid in a washing solution produced in the purification process in the wet reprocessing process is mostly recovered by a nitric acid recovery system such as evaporative concentration and reused in the wet reprocessing process, while some of the nitric acid results in excess nitric acid. This excess nitric acid is neutralized with sodium hydroxide, thereby producing a sodium nitrate liquid waste. For the neutralization reaction of the nitric acid, the nitric acid may be reacted with equimolar sodium hydroxide. Further, in the wet reprocessing process, an off-gas having nitric acid dissolved therein is produced in the dissolution step for dissolving the spent nuclear fuel in the nitric acid. The off-gas includes nitrogen oxides, a slight amount of radioactive components, and the like, and is thus washed in contact with an alkali solution with the use of a scrubber or the like in order to remove the nitrogen oxides, the radioactive component, and the like. The washing liquid waste produced in this case also includes sodium nitrate, and thus, in the present invention, this washing liquid waste is also referred to as a sodium nitrate liquid waste produced in the neutralization step with sodium hydroxide. The sodium nitrate liquid waste is then subjected to evaporative concentration to give a concentrated sodium nitrate liquid waste. The solubility of sodium nitrate in water is 91.9 g at 25° C. and 175.5 g at 100° C. with respect to 100 g of water, and a substantial amount of sodium nitrate is thus stably dissolved in the solution. Further, the melting point of sodium nitrate is 306.8° C., and sodium nitrate is decomposed at 380° C. Thus, when the sodium nitrate liquid waste is heated to approximately 100° C., water in the liquid waste is evaporated while the sodium nitrate is not moved into the water vapor but present stably in the liquid waste. Therefore, only water can be evaporated to concentrate the sodium nitrate liquid waste by heating the sodium nitrate liquid waste with an evaporator that has heating capacity to around 100° C. The water vapor generated in this case is condensed into a condensate liquid and discharged to the outside of the system. Since the concentrated sodium nitrate liquid waste thus obtained is a low-level radioactive waste, it is conventionally subjected to vitrification, cement solidification or bituminization, or kept liquid, powdered, or pelletized for intermediate storage. On the other hand, the present invention differs from prior art in that the concentrated sodium nitrate liquid waste is subjected to reductive decomposition processing with a reducing agent, and this reductive decomposition processing can achieve an advantageous effect unique to the present invention, that is, allowing for substantial decrease of waste generation due to the sodium nitrate liquid waste. As the reductive decomposition processing of the concentrated sodium nitrate liquid waste, a catalytic reduction method and a supercritical reduction method can be used which have been introduced as prior art. In each method, sodium nitrate in the sodium nitrate liquid waste is reductively decomposed to convert nitrate ions into nitrogen gas and sodium ions into sodium hydroxide and also to generate water. In addition, carbon dioxide gas is generated from the reducing agent used for the reductive decomposition. When the sodium nitrate is reductively decomposed by the catalytic reduction method, sponge copper catalysts, palladium catalysts, Raney nickel catalysts and ruthenium catalysts, and the like can be used as the catalyst, and hydrazine or a salt of hydrazine and the like can be used as the reducing agent. Alternatively, when the supercritical reduction method is employed, formic acid and the like can be used as the reducing agent. Among the products, the gas components of the nitrogen gas and carbon dioxide gas are released into the air, while the sodium hydroxide and water present in the fluid is reused as sodium hydroxide to be used for neutralization of the excess nitric acid. In addition, it is often the case that liquid wastes produced in the analyses carried out in each step of the wet reprocessing process are also nitrates, and if these liquid wastes are acidic, the wastes are subjected to neutralization processing with sodium hydroxide. Further, in the wet reprocessing process, an off-gas having nitric acid dissolved therein which is produced in the dissolution step for dissolving the spent nuclear fuel in the nitric acid includes nitrogen oxides, a slight amount of radioactive component, and the like. Thus, the off-gas is washed in contact with an alkali solution with the use of a scrubber or the like, in order to remove the nitrogen oxides, the radioactive component, and the like. Therefore, the sodium hydroxide or water produced by reductive decomposition of the sodium nitrate may be also, if necessary, recycled into the wet reprocessing process to be reused as sodium hydroxide for neutralization of liquid wastes produced in analyses in each step of the wet reprocessing process or as an alkali solution for washing an off-gas. It is to be noted that in the example described above, the sodium nitrate liquid waste produced in the neutralization step is concentrated by evaporation to give a concentrated sodium nitrate liquid waste, and the concentrated sodium nitrate liquid waste is then subjected to nitric acid reductive decomposition processing; however, the sodium nitrate liquid waste produced in the neutralization step can also be directly subjected to nitric acid reductive decomposition processing without concentration by evaporation, as indicated by an alternate long and short dash line in FIG. 1. On the other hand, since an organic solvent used as an extracting solvent in an extraction step of the wet reprocessing process is degraded under a radiation environment, the degraded component needs to be removed prior to reuse. In order to remove the degraded component, the used organic solvent is washed with sodium hydroxide, and the washing waste liquid is neutralized with nitric acid, thereby producing a sodium nitrate liquid waste. This sodium nitrate liquid waste is also subjected to concentration by evaporation in the same way as the excess nitric acid, and the obtained concentrated sodium nitrate liquid waste is reductively decomposed with a reducing agent. Among the decomposition products, the gas components of the nitrogen gas and carbon dioxide gas are released into the air, while the sodium hydroxide and water present in the fluid are reused for washing the used organic solvent. Further, if necessary, the sodium hydroxide and water are also recycled into the wet reprocessing process to be reused as sodium hydroxide for neutralization of liquid wastes produced in analyses in each step of the wet reprocessing process or as an alkali solution for washing of an off-gas. It is to be noted that also in the example described above, the used organic solvent is washed with sodium hydroxide, the sodium nitrate liquid waste produced by neutralizing the washing liquid waste with nitric acid is concentrated by evaporation to give a concentrated sodium nitrate liquid waste, and the concentrated sodium nitrate liquid waste is then subjected to nitric acid reductive decomposition processing; however, the sodium nitrate liquid waste produced by neutralizing the washing liquid waste from the organic solvent can also be directly subjected to nitric acid reductive decomposition processing without concentration by evaporation, as indicated by an alternate long and short dash line in FIG. 1. When sodium hydroxide, which is produced by reductive decomposition of the sodium nitrate liquid waste or the concentrated sodium nitrate liquid waste and recovered, is reused, radioactivity due to FP is accumulated in the sodium hydroxide with the increasing number of recycles. In that case, a part of sodium hydroxide to be recycled is brought and subjected together with other radioactive wastes to vitrification, cement solidification, bituminization, or the like for solidification, thereby allowing accumulation of radioactivity in the recycled sodium hydroxide to be reduced. In particular, when a part of the sodium hydroxide to be recycled is brought and mixed in a vitrified waste, the sodium hydroxide can be used as Na2O in borosilicate glass components, thus allowing the corresponding amount of new sodium hydroxide to be used in the neutralization step and allowing waste due to the sodium nitrate liquid waste to be cleared. A sodium nitrate solution simulating a sodium nitrate liquid waste or a concentrated sodium nitrate liquid waste is charged into a supercritical reactor, formic acid (HCOOH) is added as a reducing agent to the sodium nitrate solution, and the temperature and pressure in the reactor are then adjusted to 400° C. and 30 MPa. The conditions exceed the critical point (374° C., 22 MPa) of water, resulting in the water in a supercritical state. After keeping this state for 30 minutes, the reactor is cooled rapidly to quench the reduction reaction. The reductive decomposition products in this case include 95% of nitrogen, 4% of nitrate ions, and 1% of nitrite ions by the ratio of nitrogen, showing that most of nitrate ions in the sodium nitrate are decomposed. This reduction reaction is expressed by the following formula:2NaNO3+5HCOOH→N2+2NaOH+5CO2+4H2O This chemical reaction formula indicates the mass balance in the recycling system for the sodium salt obtained by the reductive decomposition of sodium nitrate, as stated below. More specifically, in theory, 2.0 of sodium nitrate and 1.0 of formic acid are converted to produce 1.0 of nitrogen gas, 4.0 of water, 5.0 of carbon dioxide gas and 2.0 of sodium hydroxide on a molar ratio basis. Among the products, the sodium hydroxide and water are recycled for a neutralization step of excess nitric acid generated from the wet reprocessing process, while the gas components of the nitrogen gas and carbon dioxide gas are released into the air. In addition, the recycling of the sodium hydroxide accumulates radioactivity in the sodium hydroxide. Thus, in order to reduce the accumulation of radioactivity, a part of the sodium hydroxide to be recycled is brought and charged into a vitrified high-level radioactive liquid waste. The following example shows borosilicate glass components of a vitrified waste at the reprocessing facility. SiO2: 43 to 47 wt % B2O3: 14 wt % Al2O3: 3.5 to 5 wt % Na2O: 10 wt % Other glass components: 9 to 12.5 wt % Waste oxides (excluding Na2O): 15 wt % As about 10 wt % of Na2O included in the borosilicate glass components as described above, the sodium hydroxide to be recycled can be mixed in a vitrified waste at a constant ratio, and the corresponding amount of new sodium hydroxide can be thus used in the neutralization step. As a result, a recycling process can be provided which is capable of clearing waste generation due to the sodium nitrate liquid waste. |
|
claims | 1. A system for eluting a radioactive material, comprising:an elution column, at least a portion of the elution column enclosing and being substantially filled with the radioactive material;a first sealing member sealing a first end of the elution column;a second sealing member sealing a second end of the elution column;only a single elution supply source for the system, the single elution supply source being connected to the first end of the elution column via a first fluid communication path;a collection system connected to the second end of the elution column via a second fluid communication path, the first and second fluid communication paths being in continuous fluid communication with the elution column during the eluting of the radioactive material, the first and second fluid communication paths being the only fluid communication paths fluidly connected to the elution column; anda filter on an eluant discharge end of the portion of the elution column enclosing the radioactive material, the filter being configured to support the radioactive material and prevent the radioactive material from contacting the second fluid communication path,wherein the radioactive material enclosed by the elution column is titanium Molybdate, the elution supply source introduces saline solution into the system in order to allow the saline solution to gravity drain through the elution column, and the collection system collects technetium-99m ions,wherein a length of the elution column is between about 4⅛ inches and 10⅝ inches and an inner diameter of the elution column is between about ⅝ inches and 1 3/16 inches. 2. The system of claim 1, wherein a length of the column is about 10⅝ inches and an inner diameter of the elution column is about ⅝ inches. 3. The system of claim 1, wherein a length of the column is about 5¼ inches and an inner diameter of the elution column is about 1 inch. 4. The system of claim 1, wherein a length of the column is about 4⅛ inches and an inner diameter of the elution column is about 1 and 3/16 inches. 5. The system of claim 1, wherein the filter is a glass frit and the glass frit is supported by internal shoulders of the elution column. 6. The system of claim 1, wherein the first fluid communication path is a manifold type needle including at least two subneedles. 7. The system of claim 1, wherein the portion of the elution column enclosing the radioactive material includes a plurality of undulations. 8. The system of claim 1, wherein the portion of the elution column enclosing the radioactive material is coil-shaped. 9. The system of claim 1, further comprising:a plurality of flow trippers attached to an inside wall of the portion of the elution column enclosing the radioactive material. 10. The system of claim 9, wherein each of the flow trippers have one end attached to an inside wall of the elution column and another end extending towards a center of the elution column. 11. The system of claim 9, wherein the flow trippers have a wedge shape. 12. The system of claim 9, wherein the flow trippers have an upper curved surface. 13. The system of claim 9, wherein the flow trippers are regularly arranged in a spiral pattern along a length of the column. 14. The system of claim 1, further comprising:a spiraling platform attached to an inner wall of the portion of the elution column enclosing the radioactive material. 15. The system of claim 1, wherein the portion of the elution column enclosing the radioactive material has an hourglass shape. 16. The system of claim 1, wherein the portion of the elution column enclosing the radioactive material has a funnel shape. 17. The system of claim 1, wherein the system does not require any adsorbent material on a downstream side of the filter on the eluant discharge end of the portion of the elution column enclosing the radioactive material. 18. The system of claim 1, wherein the system does not require adsorbent packing material. 19. A system for eluting a radioactive material, comprising:an elution column, at least a portion of the elution column enclosing and being substantially filled with the radioactive material;a first sealing member sealing a first end of the elution column;a second sealing member sealing a second end of the elution column;an elution supply source for the system, the single elution supply source being connected to the first end of the elution column via a first fluid communication path, the system not including more than one elution supply source;a collection system connected to the second end of the elution column via a second fluid communication path, the first and second fluid communication paths being in continuous fluid communication with the elution column during the eluting of the radioactive material, the first and second communication paths being the only fluid communication paths fluidly connected to the elution column; anda filter on an eluant discharge end of the portion of the elution column enclosing the radioactive material, the filter being configured to support the radioactive material and prevent the radioactive material from contacting the second fluid communication path,wherein the radioactive material enclosed by the elution column is titanium Molybdate, the elution supply source introduces saline solution into the system in order to allow the saline solution to gravity drain through the elution column, and the collection system collects technetium-99m ions,wherein a length of the elution column is between about 4⅛ inches and 10⅝ inches and an inner diameter of the elution column is between about ⅝ inches and 1 3/16 inches. 20. The system of claim 19, wherein the system does not require any adsorbent material on a downstream side of the filter on the eluant discharge end of the portion of the elution column enclosing the radioactive material. 21. The system of claim 19, wherein the system does not require adsorbent packing material. 22. The system of claim 19, wherein,the first sealing member is part of the first end of the elution column,the second sealing member is part of the second end of the elution column,the elution column is a uniform diameter from the first end of the elution column to the second end of the elution column. 23. A system for eluting a radioactive material, comprising:an elution column with a first end and a second end, at least a portion of the elution column enclosing and being substantially filled with the radioactive material;a first sealing member included in the first end of the elution column, the first sealing member being configured to seal the first end of the elution column;a second sealing member included in the second end of the elution column, the second sealing member being configured to seal the second end of the elution column;only a single elution supply source for the system, the single elution supply source being connected to the first end of the elution column via a first fluid communication path;a collection system connected to the second end of the elution column via a second fluid communication path, the first and second fluid communication paths being in continuous fluid communication with the elution column during the eluting of the radioactive material, the first and second fluid communication paths being the only fluid communication paths fluidly connected to the elution column; anda filter on an eluant discharge end of the portion of the elution column enclosing the radioactive material, the filter being configured to support the radioactive material and prevent the radioactive material from contacting the second fluid communication path,wherein the elution column is a uniform diameter from the first end of the elution column to the second end of the elution column. 24. The system of claim 23, wherein,the system is configured to allow an elution solution to gravity drain through the elution column,a length of the elution column is between about 4⅛ inches and 10⅝ inches and an inner diameter of the elution column is between about ⅝ inches and 1 3/16 inches. |
|
047023910 | abstract | A containment with a long-time corrosion resistant protective cover for a container with highly radioactive content consists of a thick-walled metallic containment cylinder having a bottom and lid mounted at its bottom and upper ends completely surrounded and sealed in by corrosion protective linings and layers to prevent corrosion damage. Additionally, cover and bottom plates are mounted on the lid and the bottom, respectively, which plates have diameters larger than those of the containment cylinder and the lid and bottom in order to prevent mechanical damage to the corrosion protective layers and liners of the containment. |
051280680 | claims | 1. A method of treating particulate material contaminated with heavy metal compounds, radioactive compounds, and organics, single and in combination, said method comprising the steps of: washing said particulate material with a first quantity of contaminant mobilizing solution selected from a group consisting of leaching solutions, surfactants and mixtures thereof to mobilize soluble and dispersible contaminants in the liquid phase of said contaminate mobilizing solution; mechanically separating large particles of a size substantially free of residual contamination from intermediate sized particles, fines, and said contaminant mobilizing solution; washing the separated large particles with a water based cleaning solution to produce recovered particulate material; size separating fines from said intermediate sized particles with a countercurrent flow of additional contaminant mobilizing solution to form a waste slurry, said size separating of said fines by countercurrent flow being performed by introducing the intermediate sized particles, fines, and said first quantity of contaminant mobilizing solution into a size separator in a flow with a first flow direction while introducing additional contaminant mobilizing solution in a countercurrent flow with a second flow direction counter to said first flow direction whereby the fines are carried with said counterflow as part of the waste slurry; attrition abrading the intermediate sized particles from which fines have been removed by said countercurrent flow of contaminant mobilizing solution to dislodge attached fines from their intermediate sized particles; size separating the dislodged fines from the intermediate sized particles by a countercurrent flow of wash water to form additional waste slurry and effluent of said intermediate sized particles and wash water, said size separating of said dislodged fines by countercurrent flow being performed by introducing the intermediate particles and dislodged fines into a size separator in a flow with a first flow direction while introducing wash water in a countercurrent flow in a second flow direction counter to said first flow direction whereby the dislodged fines are carried with said counterflow as part of the additional waste slurry and effluent; and dewatering the effluent of intermediate sized particles and wash water to produce additional recovered particulate materials. washing said particulate material with a contaminant mobilizing solution selected from a group consisting of leaching solutions, surfactants and mixtures thereof to mobilize soluble and dispersible contaminants in the liquid phase of said contaminate mobilizing solution; mechanically separating large particles of a size substantially free of residual contamination from intermediate sized particles, fines, and said contaminant mobilizing solution; washing the separated large particles with a water based cleaning solution to produce recovered particulate material; size separating said fines from said intermediate sized particles by a countercurrent flow of wash water to form a waste slurry and an effluent of intermediate size particles and wash water, said size separating of said fines by counter-current flow being performed by introducing the intermediate particles and fines into a size separator in a flow with a first flow direction while introducing wash water in a countercurrent flow in a second flow direction counter to said first flow direction whereby the dislodged fines are carried with said counterflow as part of the waste slurry and effluent; density separating heavy metal particles from said effluent of intermediate sized particles and wash water; and dewatering the effluent of intermediate sized particles and wash water to produce additional recovered particulate materials. introducing said slurry in a first flow direction into the top of said mineral jig; introducing a countercurrent flow of fluid in said mineral jig in a second flow direction counter to said first flow direction; and withdrawing an underflow from the mineral jig at a rate, related to the countercurrent flow of fluid and the introduction of said slurry, which produces an overflow of fluid containing said fines and an underflow containing said larger particles of said particulate material. 2. The method of claim 1 including attrition abrading the intermediate sized particles to initially dislodge attached fines prior to size separating said fines from the intermediate sized particles with the countercurrent flow of said contaminant mobilizing solution. 3. The method of claim 2 including density separating heavy metal compounds from said effluent of intermediate sized particles and wash water by a cross-current flow of wash water prior to dewatering said effluent of intermediate sized particles and wash water. 4. The method of claim 1 including density separating heavy metal compounds from said effluent of intermediate sized particles and wash water prior to dewatering. 5. The method of claim 4 wherein said density separating of heavy metal compounds from the effluent of intermediate sized particles in wash water is effected by a cross-current flow of wash water. 6. The method of claim 5 wherein water removed from said effluent of intermediate sized particles and wash water by dewatering is recycled as said wash water for said size separating with the countercurrent, flow of wash water. 7. The method of claim 1 wherein water removed from said effluent of intermediate sized particles and wash water by dewatering is recycled as said wash water for said size separating with the countercurrent flow of wash water. 8. The method of claim 7 wherein the dewatered intermediate size particles are further dried by evaporation. 9. The method of claim 1 including separating said fines from said contaminant mobilizing solution in said waste slurry and additional waste slurry and recycling said contaminant mobilizing solution. 10. The method of claim 9 including removing organics from the recycled contaminant mobilizing solution. 11. The method of claim 9 including removing radioactive contaminants from said recycled solution. 12. The method of claim 11 including removing organics from said recycled solution. 13. The method of claim 1 wherein large particle greater than about 5 mm in size are mechanically separated from the intermediate sized particles and fines. 14. The method of claim 13 wherein the countercurrent flows of said contaminant mobilizing solution and said wash water are set to remove fines of a size selected to reduce contamination to a predetermined level. 15. The method of claim 14 wherein said countercurrent flows of contaminant mobilizing solution and wash water are selected to separate fined of a size smaller than about 200 microns. 16. The method of claim 15 wherein said countercurrent flows of contaminant mobilizing solution and wash water are selected to separate fined of a size smaller than about 60 microns. 17. The method of claim 1 wherein said separating steps are carried out by operating a mineral jig in a countercurrent flow mode. 18. A method of treating particulate material contaminated with heavy metal compounds, radioactive compounds, and organics, single and in combination, said method comprising the steps of: 19. The method of claim 18 wherein said density separating of heavy metals from the effluent of intermediate sized particles in wash water is effected by a cross-current flow of wash water. 20. The method of claim 18 wherein water removed from said effluent of intermediate sized particles and wash water by dewatering is recycled as said wash water for said size separating with the countercurrent flow of wash water. 21. A method of operating a mineral jig to separate fines from larger particulates in a slurry of particulate material and to simultaneously wash sad particulate material comprising: 22. The method of claim 21 wherein said countercurrent flow is introduced at a rate of about 1 to 8 liters/min and said underflow is withdrawn at a rate of about 1 to 3 liters/min. 23. The method of claim 21 wherein said particulate matter selectively contains soluble and dispersible contaminants and said countercurrent flow is a flow of a contaminant mobilizing solution which mobilizes said soluble and dispersible contaminants and carries them off in the overflow with said fines. |
abstract | A computer implemented system and method for providing general data protection regulation (GDPR) compliant hashing in blockchain ledgers. The invention guarantees a user's right to be forgotten, in compliance with the GDPR regulations, utilizing blockchain technologies. |
|
claims | 1. A radiation therapy apparatus comprising:a gantry comprising a central aperture that defines a cavity, in which a patient is positionable for irradiation;a positioning device, with which the patient is positionable in the central aperture of the gantry;a radiation head for generating therapeutic radiation that is directable at the patient; andan extension mechanism for the radiation head,wherein the extension mechanism is configured such that, in a first operating mode of the radiation therapy apparatus, the radiation head is disposed inside the gantry such that the gantry has, on a side facing the positioning device, a front surface that is a front limit of the radiation therapy apparatus toward the positioning device, and that in the first operating mode, the radiation head is behind the front surface viewed from the positioning device, andwherein the extension mechanism is configured such that, in a second operating mode of the radiation therapy apparatus, the radiation head, viewed from the positioning device, is movable to be in front of the front surface that in the first operating state is the front limit. 2. The radiation therapy apparatus as claimed in claim 1, wherein the gantry has, on a side facing away from the positioning device, a back surface that, in the first operating mode, is a rear limit of the gantry, andwherein the extension mechanism is configured, such that in a third operating mode of the radiation therapy apparatus, the radiation head is movable, using the extension mechanism, to be behind the back surface viewed from the positioning device. 3. The radiation therapy apparatus as claimed in claim 2, wherein the gantry is tiltable about a horizontal axis. 4. The radiation therapy apparatus as claimed in claim 2, wherein the central aperture of the gantry has a longitudinal axis, andwherein the extension mechanism enables a translatory movement of the radiation head along a direction having a component along the longitudinal axis. 5. The radiation therapy apparatus as claimed in claim 2, wherein the extension mechanism is extendable out along an isocentric circular segment. 6. The radiation therapy apparatus as claimed in claim 1, wherein the gantry is tiltable about a horizontal axis. 7. The radiation therapy apparatus as claimed in claim 6, wherein the central aperture of the gantry has a longitudinal axis, andwherein the extension mechanism enables a translatory movement of the radiation head along a direction having a component along the longitudinal axis. 8. The radiation therapy apparatus as claimed in claim 1, wherein the central aperture of the gantry has a longitudinal axis, andwherein the extension mechanism enables a translatory movement of the radiation head along a direction having a component along the longitudinal axis. 9. The radiation therapy apparatus as claimed in claim 8, wherein the extension mechanism is configured such that, in addition to the translatory movement, a tilting movement of the radiation head that tilts the radiation head relative to the extension mechanism is executable. 10. The radiation therapy apparatus as claimed in claim 8, wherein the extension mechanism provides only a translatory movement of the radiation head in the direction of the longitudinal axis. 11. The radiation therapy apparatus as claimed in claim 10, wherein the extension mechanism is configured such that, in addition to the translatory movement, another movement of the radiation head is executable, the other movement having a component in a direction perpendicular to the translatory movement. 12. The radiation therapy apparatus as claimed in claim 10, wherein the extension mechanism is configured such that, in addition to the translatory movement, a tilting movement of the radiation head that tilts the radiation head relative to the extension mechanism is executable. 13. The radiation therapy apparatus as claimed in claim 10, wherein the extension mechanism is extendable out along an isocentric circular segment. 14. The radiation therapy apparatus as claimed in claim 8, wherein the extension mechanism is configured such that, in addition to the translatory movement, another movement of the radiation head is executable, the other movement having a component in a direction perpendicular to the translatory movement. 15. The radiation therapy apparatus as claimed in claim 14, wherein the extension mechanism is configured such that, in addition to the translatory movement, a tilting movement of the radiation head that tilts the radiation head relative to the extension mechanism is executable. 16. The radiation therapy apparatus as claimed in claim 14, wherein the extension mechanism is configured such that the radiation head is alignable to an isocentric emission of a beam of the therapeutic radiation. 17. The radiation therapy apparatus as claimed in claim 6, wherein the extension mechanism is extendable out along an isocentric circular segment. 18. The radiation therapy apparatus as claimed in claim 8, wherein the extension mechanism is configured such that the radiation head is alignable to an isocentric emission of a beam of the therapeutic radiation. 19. The radiation therapy apparatus as claimed in claim 1, wherein the extension mechanism is extendable out along an isocentric circular segment. 20. The radiation therapy apparatus as claimed in claim 1, wherein the extension mechanism is configured such that the radiation head is alignable to an isocentric emission of a beam of the therapeutic radiation. |
|
046876260 | summary | BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention is directed to nuclear power reactors. More specifically, this invention is related to a safety device for nuclear power reactors. 2. Description of the Prior Art In a conventional nuclear power reactor system, the reactor and the steam generator connected thereto are situated within a containment wall. The steam so generated is fed via a pipe through the containment wall to drive a steam turbine for power generation. Typically, to guard against over pressurization within the pipe, steam safety valves and steam release valves are put on line. Thus, in the event of a rupture in the pipe, the main steam isolation valve is closed and high pressure steam is released or dumped into the surrounding atmosphere through the steam release valves. Such a release into the atmosphere runs the risk of releasing radioactive materials into the air, particularly if some of the tubes in the steam generator in the nuclear reactor should rupture simultaneously. Hence, there is a need for a passive emergency steam dumping system to be used in connection with nuclear reactors. By passive, it is meant devices which will perform its assigned duty even when electric power is not available. Furthermore, when there is a total loss of electric power coupled with a steam line break, removal of the residual heat in the steam generator and the emergency auxiliary feedwater system will need an uninterrupted power source to circulate cold water to condense the steam and to supply the auxiliary feed water. Although emergency generators can be used as a back-up in case of power failure, it would be highly preferable to have a passive system whereby the energy in the high pressure steam from the steam generator is converted into mechanical energy for heat removal and circulation of emergency feedwater. Since passive systems do not require an energy source as a driving force, the presence of such a passive system for use during power failures is clearly desirable. SUMMARY OF THE INVENTION According to the present invention, a passive emergency steam dump for nuclear power reactors is provided which comprises: a first storage means for coolant water; PA1 a steam jet ejector disposed in the first storage means, the operating steam for the ejector being supplied by the steam generator; PA1 a heat exchange means having its inlet connected to the outlet of the steam jet ejector and the outlet, to the first storage means; and PA1 a second storage means for coolant water in which the heat exchanger is immersed. In another embodiment of the subject invention, the outlet of the steam jet ejector is also connected to the main feedwater line for the steam generator for returning condensate from the steam jet ejector to the steam generator when the pressure in the outlet exceeds a specified limit. |
062366998 | claims | 1. In a pressurized water nuclear power plant having a nuclear reactor core and a multiplicity of control rods arranged as groups for movement sequentially through the core for controlling the reactor power while exposed to the nuclear reactions in the core, a system for the operator to monitor compliance with administrative limits on the cumulative exposure in the core of the control rod groups, comprising: means for measuring the core power and generating a power signal commensurate therewith; means for measuring the position of each control rod group in the core; means for establishing an incremental time base common to the measuring of the core power and the measuring of the position of each group; means responsive to said means for measuring of position, for determining when on said time base, each group is within said position range; means for determining the core power when each group is in said position range; means for computing an incremental effective exposure for each group, commensurate with core power, for each time increment at which each group is within said position range; means for accumulating said incremental effective exposure for each group; means for comparing the accumulated effective exposure for each group with the administrative limit for each group; and means for displaying said comparison to the reactor operator. said means for displaying include at least one scale of uniform intervals, marked with a plurality of numeric values indicative of an initial zero value and a final valve corresponding to the reference period; an indicator configuration for each group, each indicator configuration having a scale associated therewith, and consisting of an indicator for each component of actual exposure, each indicator initially appearing at the zero representation of the scale and growing in size toward said final value to span the number of scale intervals corresponding to the ratio of effective exposure of the component to the reference period; means for advancing each indicator along said scale toward the said final value, at a uniform rate, independently of but simultaneously with said indicator growth. each of said scales is displayed for a respective group, in the form of a circle with coincident zero and said final value; and each display of an indicator of a component is a sector of the circle, said sector having a variable angle and which advances by continually rotating about the center of the circle toward said final value. one scale is a linear segment with the zero value at one end and said final value at the other end; the indicator configuration for each of at least two groups is associated with said one scale; and each indicator of a component is displayed as a horizontal bar, which grows by increasing in horizontal length and which advances by continually moving horizontally toward said final value. 2. The system of claim 1, wherein said administrative limit is in the form of a limit index defined by a maximum permissible hours of accumulated effective exposure during a reference period of exposure; and 3. The system of claim 2, wherein 4. The system of claim 2, wherein |
claims | 1. A computer-implemented diagnostic method for determining a cause of an event in a product or process according to an energy function model, the method comprising:using at least one processor to perform the steps of:receiving a signal from a first measurement system indicating the event in the product or process;displaying, on a display device, a schematic of a function model of the product or process, the function model comprising a plurality of functions describing actions designed to be performed in the operation of the product or process, the function model further identifying dependencies between the plurality of functions to describe how, why, and when respective actions are performed;converting the function model to an energy function model by performing, using the at least one processor, the steps of:(a) assigning, to each of the plurality of functions, an energy function according to an energy action describing how the particular function of the product or process uses energy during operation;(b) identifying, for a first energy function, a plurality of energy paths within the product or process;(c) displaying, on the schematic, the energy functions and a representation of the plurality of energy paths shown with respect to a displayed how direction, an opposing why direction, and a perpendicular when direction corresponding to the dependencies between the plurality of functions; and(d) associating the first energy function with a second energy function by displaying, on the schematic, a connection between at least one of the plurality of energy paths for the first energy function and the second energy function according to an energy relationship between the first energy function and the second energy function to define the energy function model so as to account for energy within the product or process during operation;selecting, from among the plurality of energy paths, a measurement energy for detecting a contrast between how the product or process is actually using energy and how the product or process is intended to use energy;generating a plurality of measurements of the selected measurement energy using a second measurement system that is different than the first measurement system that indicated the event in the product or process;displaying the plurality of measurements on the display device to identify the contrast; andconducting a progressive search on the contrast to identify a feature or property of the product or process responsible for causing the event. 2. The diagnostic method according to claim 1, wherein the event comprises a malfunction event. 3. The diagnostic method according to claim 1 further comprising the step of limiting the schematic to the narrowest scope known to contain the root cause of the event. 4. The diagnostic method according to claim 1, wherein the first energy function comprises a function for directing energy. 5. The diagnostic method according to claim 1, wherein the first energy function comprises a function for transmitting energy. 6. The diagnostic method according to claim 1, wherein the first energy function comprises a function for converting energy. 7. The diagnostic method according to claim 1, wherein the first energy function comprises a function for containing energy. 8. The diagnostic method according to claim 1, wherein at least one energy path of the plurality of energy paths comprises an input energy path corresponding to energy used to achieve the assigned first energy function. 9. The diagnostic method according to claim 1, wherein at least one energy path of the plurality of energy paths comprises an output energy path corresponding to the performance of useful work. 10. The diagnostic method according to claim 1, wherein at least one energy path of the plurality of energy paths comprises a waste energy path corresponding to energy loss. 11. The diagnostic method according to claim 1, wherein at least one energy path of the plurality of energy paths comprises an input signal energy path. 12. The diagnostic method according to claim 1, wherein at least one energy path of the plurality of energy paths comprises an environmental energy path. 13. The diagnostic method according to claim 1, wherein the plurality of generated measurements comprises direct measurements of the selected energy. 14. The diagnostic method according to claim 1, wherein the plurality of generated measurements comprises indirect measurements of the selected energy through at least one of its component factors. 15. The diagnostic method according to claim 1 further comprising the step of controlling the identified feature or property to prevent a future occurrence of the event. 16. The diagnostic method according to claim 2, wherein the contrast between how the product or process is actually performing and how the product or process is intended to perform is detected by generating a plurality of energy measurements for a second product or process that is not experiencing a malfunction event. 17. The diagnostic method according to claim 2, wherein the product or process comprises a prototype product or process. 18. The diagnostic method according to claim 17, wherein the identified feature or property corresponds to a design under consideration for which a contrast in the direct measurement of the malfunction event is not detected. 19. The diagnostic method according to claim 1, wherein the product or process comprises a production product or process. 20. The diagnostic method according to claim 1 further comprising the step of storing the plurality of generated measurements in a computer data storage device. 21. The diagnostic method according to claim 1, wherein selecting the measurement energy comprises:selecting a first energy;generating one or more first measurements of the selected first energy in the product or process;based on the one or more first measurements, determining that the contrast is not detected by measuring the first energy;in response to determining that the contrast is not detected by measuring the first energy, selecting a second energy;generating one or more second measurements of the selected second energy in the product or process;based on the one or more second measurements, determining that the contrast is detected by measuring the second energy; andassigning the second energy as the measurement energy. 22. A computer-implemented method for diagnosing and applying corrective action to a malfunctioning product or process according to an energy function model, the method comprising:using one or more processors to perform the steps of:receiving a signal from a first measurement system indicating the event in the product or process;selecting from among a plurality of energy paths existing during operation of the product or process, a first energy for measurement to detect a contrast between how the product or process is actually using energy and how the product or process is intended to use energy, wherein the energy function model defines the plurality of energy paths associated with respective energy functions of the product or process, wherein the respective energy functions describe how the product or process is intended to use energy during operation, and wherein the one or more processors displays on a display device a representation of at least one of the plurality of energy paths connecting two or more of the respective energy functions along a displayed how direction, an opposing why direction, and a perpendicular when direction corresponding to dependencies between the two or more respective functions so as to account for energy within the product or process during operation;conducting a progressive search on the contrast to identify a feature or property of the product or process that can be used to control an energy function that is malfunctioning; andcontrolling the identified feature or property to prevent future malfunctions. 23. The method according to claim 22, wherein the contrast between how the product or process is actually performing and how the product or process is intended to perform is detected by generating a plurality of energy measurements for a second product or process that is not experiencing a malfunction event. 24. The method according to claim 22, wherein the product or process comprises a prototype product or process. 25. The method according to claim 24, wherein the identified feature or property corresponds to a design under consideration for which a contrast in the direct measurement of the malfunction event is not detected. 26. The method according to claim 22, wherein the product or process comprises a production product or process. 27. A non-transitory computer-readable medium having stored thereon computer-executable program instructions for diagnosing a cause of an event in a product or process according to an energy function model, the computer-executable program instructions, when executed by a computer processor, causing the computer processor to perform the steps of:receiving a signal from a first measurement system indicating the event in the product or process;displaying, on a display device, a schematic of a function model of the product or process, the function model comprising a plurality of functions describing actions designed to be performed in the operation of the product or process, the function model further identifying dependencies between the plurality of functions to describe how, why, and when respective actions are performed;converting the function model to an energy function model by performing, using the computer processor, the steps of:(a) assigning, to each of the plurality of functions, an energy function according to an energy action describing how the particular function of the product or process uses energy during operation;(b) identifying, for a first energy function, a plurality of energy paths within the product or process;(c) displaying, on the schematic, the energy functions and a representation of the plurality of energy paths shown with respect to a displayed how direction, an opposing why direction, and a perpendicular when direction corresponding to the dependencies between the plurality of functions; and(d) associating the first energy function with a second energy function by displaying, on the schematic, a connection between at least one of the plurality of energy paths for the first energy function and the second energy function according to an energy relationship between the first energy function and the second energy function to define the energy function model so as to account for energy within the product or process during operation;selecting, from among the plurality of energy paths, a measurement energy for detecting a contrast between how the product or process is actually using energy and how the product or process is intended to use energy;generating a plurality of measurements of the selected measurement energy using a second measurement system that is different than the first measurement system that indicated the event in the product or process; andconducting a progressive search on the contrast to identify a feature or property of the product or process responsible for causing the event. |
|
description | 1. Field of the Invention The present invention relates to a particle beam rotational irradiation apparatus (a rotating gantry), the objective of which is to irradiate a charged particle, accelerated by an accelerator, from an arbitrary angle direction. 2. Description of the Related Art Charged particles are circulated and accelerated in a circular accelerator such as a synchrotron and then the charged particles (mainly protons or carbon ions), which have been accelerated to gain high energy, are extracted from the circulation orbit; the charged particles (referred to also as charged particle beam or particle beam) are utilized in a physics experiment or in a particle beam therapy such as a cancer treatment, in which the charged particles are transported through a beam transport line and irradiated onto a desired subject. In general, in a cancer treatment utilizing accelerated charged particles, i.e., in a so-called particle beam therapy, the irradiation directions are changed for the purpose of averting the charged particles from major organs during the treatment or preventing normal tissues from being damaged. As one of the means of changing the irradiation directions of the charged particle beams, a particle beam rotational irradiation apparatus (rotating gantry) is commonly utilized in which an irradiation nozzle is mounted in a structural member that rotates around a patient and the charged particles can be irradiated from a desired angle. For example, in a rotating gantry (referred to simply as a gantry, as may be necessary) disclosed in Patent Document 1, a charged particle beam is once bended by two bending electromagnets in such a way that the beam transport line becomes perpendicular to the rotation axis of the rotating gantry; then, the charged particle beam is introduced by bending the charged particle beam again by use of two bending electromagnets in such a way that on a plane perpendicular to the center axis of the gantry, the charged particle beam is directed to the isocenter (the intersection point of the gantry rotation axis with the beam axis, which is the reference of the irradiation target). This kind of beam transport line makes the length of the rotating gantry shortest with respect to the rotation-axis direction; therefore, as a result, it is made possible to install a rotating gantry in a small area. In some cases, a rotating gantry provided with this kind of beam transport line is referred to as a corkscrew-type gantry. In addition, the bending electromagnets are two-pole electromagnets provided with two magnetic poles. It is another characteristic of a corkscrew-type gantry that at least four bending electromagnets are required. The reason for that will be explained below. In general, the momentum of a charged particle beam extracted from a circular accelerator does not have a certain single value but has spread around the center value. The value obtained through dividing a deviation amount from the center value by the center value is referred to as a momentum spread. When a charged particle beam having a momentum spread passes through the bending electromagnet, the bending angle thereof changes depending on the momentum (it might be considered as energy or a velocity) of the charged particle beam; therefore, when the spread is left as it is, the width (referred to as a beam size, hereinafter) of particle distribution at the isocenter may become large. Letting p0 and Δp/p0 denote the momentum (center momentum) of a particle having the center momentum at a given position and the momentum spread, respectively, the deviation (the spread of a beam width due to the momentum spread) Δx from the center orbit, which is the orbit of the particle having the center momentum p0, is given as the equation (1) by use of a dispersion function η that characterized the effect of the momentum spread at the given position. The dispersion function η is a function of the position of a beam transport line.Δx=η×Δp/p0 (1) In general, as an element for causing the dispersion function η, a bending electromagnet is utilized; when the dispersion function η once becomes a value other than “0”, it is required to cancel η and η′ by use of at least another bending electromagnet and quadrupole electromagnets. Here, η′ denotes the differentiation in the beam traveling direction (s direction, s axis). Because an actual irradiation site is not a point like an isocenter and has a width in the depth direction, it is required to nullify the gradient η′ of the dispersion function η. A dynamic change of a momentum spread makes the charged particle beam look like moving. In general, it is required, in a rotating gantry, that the dispersion function η is diminished at the isocenter to the extent that its contribution to the spread of a beam width can be allowed, in order to prevent the beam width from changing or moving at the isocenter. In the case of the corkscrew-type gantry disclosed in Patent Document 1, two bending electromagnets at the upstream side are situated on the same plane; therefore, the dispersion function η caused by the first bending electromagnet is nullified by the other bending electromagnet that has the bending plane thereof on the same plane. In this situation, a plurality of quadrupole electromagnets provided between the bending electromagnets are utilized for changing the s-direction gradient (η′) of the dispersion function η in addition to focusing or defocusing a charged particle beam. The bending planes of two bending electromagnets at the downstream side differ by 90 degrees from those of the two bending electromagnets at the upstream side; as is the case with the upstream side, η and η′ are nullified by the two electromagnets and a plurality of quadrupole electromagnets provided between the bending electromagnets. In general, in the designing of the beam transport line between a circular accelerator and a rotating gantry or in the designing of the beam transport line of a rotation gantry, there exists a type in which all bending electromagnets are arranged in a single bending direction and nullify η and η′, which are caused in only a single direction, in collaboration with quadrupole electromagnets, or a type, as disclosed in Patent Document 1, in which a plurality of bending electromagnets are arranged in such a way that although respective η and η′ are caused in both the x direction and the y direction, 90-degree-different bending planes makes the x-direction dispersion function η and the y-direction dispersion function η not couple with each other, i.e., the x-direction dispersion function η and the y-direction dispersion function η are independent in the respective directions thereof. Next, a gantry in which η and η′ are caused in only a single direction will be explained. In the rotating gantry disclosed in Patent Document 2, three bending electromagnets lead a charged particle beam to the isocenter. Because three bending electromagnets are provided, the dispersion function is nullified in only one direction in the designing of the beam transport line of the rotating gantry, unless the coupling is not utilized; thus, the bending planes of a charged particle beam is made to be a single and the same by the three bending electromagnets. Accordingly, the beam transport line of the rotating gantry disclosed in Patent Document 2 becomes longer in the beam-rotation-axis direction than that of the corkscrew-type gantry; as a result, the area where the rotating gantry is installed becomes wider. In recent years, it has been required to raise the throughput of particle beam therapy, as particle beam therapy has become widespread. In particle beam therapy, before a charged particle beam is irradiated onto a patient, a person who offers assistance in the therapy, for example, a radiologist approaches a patient platform or an irradiation nozzle and fixes the patient body or adjusts irradiation-system apparatuses that are to be mounted on an irradiation nozzle. In this situation, in order to shorten the time for the adjustment, i.e., in order to raise the throughput of the therapy, the easiness degree of the foregoing work at the vicinity of the irradiation nozzle is important. An example of rotating gantry that facilitates the work at the vicinity of the irradiation nozzle is disclosed, for example, in Patent Document 3. By constructing a rotating gantry (referred to as an open type, hereinafter) in such a way that as illustrated in FIG. 6 of Patent Document 3, the front end of the irradiation apparatus (irradiation nozzle) protrudes toward the irradiation room, the work to be performed in the vicinity of the irradiation nozzle can be facilitated. Moving body tracking, in which high-accuracy irradiation is performed while the motion of an organ undergoing irradiation or the motion of a body caused by respiration is monitored in real time, has also been attracting attention; therefore, it is desired to secure a sufficient space for arranging monitoring apparatuses in the vicinity of the irradiation nozzle. [Patent Document 1] U.S. Pat. No. 4,917,344 (FIGS. 1a and 1b) [Patent Document 2] International Publication No. WO2008/026648A1 (FIG. 1) [Patent Document 3] Japanese Patent Application Laid-Open No. 2006-192297 (FIG. 6) [Patent Document 4] Japanese Patent Application Laid-Open No. 2000-140134 (FIGS. 1, 2, and 6) As the rotating gantry disclosed in Patent Document 2 can be realized to construct, the open-type rotating gantry disclosed in Patent Document 3 can be realized to construct, because three bending electromagnets lead a charged particle beam to the isocenter. However, the rotating gantry (corkscrew-type gantry) disclosed in Patent Document 1 is a type that requires a small installation area; thus, it is difficult to apply an open-type structure to this kind of rotating gantry. The reason for that will be explained with reference to the drawings of Patent Document 4. The rotating gantry disclosed in Patent Document 4 is a corkscrew-type gantry, as is the case in Patent Document 1. FIGS. 1 and 2 of Patent Document 4 illustrate a left-side view and an elevation view of a rotating gantry, respectively; FIG. 6 of Patent Document 4 illustrates a left-side view of a rotating frame in which an isocenter CI is shown. The rotating gantry disclosed in Patent Document 4 is a type in which the irradiation nozzle (Reference Numeral 15 of Patent Document 4) is provided inside the rotating frame (Reference Numeral 2 of Patent Document 4) and the isocenter CI is situated in the rotating frame. The rotating gantry disclosed in Patent Document 4 is different from a rotating gantry in which as is the case in Patent Document 3, the irradiation nozzle (Reference Numeral 8 of Patent Document 3) is provided outside the rotating frame (Reference Numeral 1 of Patent Document 3). In the corkscrew-type gantry disclosed in Patent Document 1 or 4, two bending electromagnets at the downstream side are arranged in a row on a plane perpendicular to the rotation axis so that the installation area is diminished; therefore, in order to realize an open-type rotating gantry in which an irradiation nozzle to be disposed under the last bending electromagnet is provided in the treatment room, not only the irradiation nozzle and the downstreammost bending electromagnet but also another bending electromagnet, which forms a pair with the downstreammost bending electromagnet, are made to protrude into the treatment room. Because it is difficult to support the two bending electromagnets and the irradiation nozzle by the frame of the rotating gantry, it is difficult to realize a corkscrew-type gantry having an open-type rotating gantry. Accordingly, as is the case in the rotating gantry disclosed in Patent Document 2, in the corkscrew-type gantry disclose in Patent Document 1 or 4, it is inevitable to perform irradiation onto a patient in a hole-shaped narrow space (inner chamber); therefore, the work to be performed in the vicinity of the irradiation nozzle becomes hard. Moreover, even if the two bending electromagnets and the irradiation nozzle can be arranged in the treatment room, the space for disposing monitoring apparatuses around the irradiation nozzle or the space that facilitates the work to be performed in the vicinity of the irradiation nozzle cannot sufficiently be secured. To date, such a corkscrew-type gantry as disclosed in Patent Document 1 has been utilized in order to cancel the momentum-spread dependence of a beam position at the isocenter. In order to realize an open-type rotating gantry, it is required that as is the case with a rotating gantry that requires a large installation area, i.e., as is the case with the rotating gantry disclosed in Patent Document 2 or 3, the bending electromagnets are arranged in such a way that the bending planes of all the bending electromagnets are the same; therefore, there has been a problem that it is difficult to realize a gantry that requires a small installation area. In an open-type rotating gantry, as disclosed in Patent Document 3, that is suitable to raise the throughput of particle beam therapy, the irradiation nozzle is made to protrude into the irradiation room; therefore, the patient-positioning work performed in the vicinity of the irradiation nozzle can be facilitated. However, because the bending electromagnets are arranged in such a way that the bending planes of all the bending electromagnets are the same, there has been a problem that the installation area becomes large. The objective of the present invention is to realize a particle beam rotational irradiation apparatus that can raise the throughput of particle beam therapy and can be downsized. A particle beam rotational irradiation apparatus according to the present invention is provided with an irradiation nozzle that irradiates a charged particle beam, a beam transport unit that transports the charged particle beam to the irradiation nozzle, and a rotating unit that can rotate around the isocenter; the particle beam rotational irradiation apparatus is characterized in that the beam transport unit has three or more bending electromagnets and in that the bending electromagnets are arranged in such a way that in the case where as a pair of bending planes, any two of the bending planes of the bending electromagnets are selected, the two bending planes of at least one pair of bending planes are not on the same plane, not parallel with each other, and not perpendicular to each other. In a particle beam rotational irradiation apparatus according to the present invention, the bending electromagnets are arranged in such a way that in the case where as a pair of bending planes, any two of the bending planes of the bending electromagnets are selected, the two bending planes of at least one pair of bending planes are not on the same plane, not parallel with each other, and not perpendicular to each other; therefore, the working space for performing patient positioning work can be widened even in the case of a small-size particle beam rotational irradiation apparatus and hence the throughput of particle beam therapy can be improved. The foregoing and other object, features, aspects, and advantages of the present invention will become more apparent from the following detailed description of the present invention when taken in conjunction with the accompanying drawings. FIG. 1 is a side view illustrating a particle beam rotational irradiation apparatus according to Embodiment 1 of the present invention; FIG. 2 is an elevation view illustrating a particle beam rotational irradiation apparatus according to Embodiment 1 of the present invention. FIG. 3 is a view illustrating a beam transport unit according to Embodiment 1 of the present invention. A particle beam rotational irradiation apparatus 20 is provided with a rotating unit 1 that rotates around the isocenter, supporting bases 4a and 4b that support the rotating unit 1, and a rotation-driving system (unillustrated) that makes the rotating unit 1 rotate. The rotating unit 1 is provided with a body unit 13, a beam transport unit 15 that transports a charged particle beam 14, an irradiation nozzle 8 (refer to FIG. 3) that irradiates the charged particle beam 14 onto an irradiation subject 19, and a weight 7 whose weight is balanced with that of the beam transport unit 15. The body unit 13 has a front ring 2, a rear ring (bearing) 3, and a plurality of supporting members 10; the body unit 13 is a structure that supports the beam transport unit 15. FIG. 3 is a view of the beam transport unit 15 and the irradiation nozzle 8 taken for the sake of easier understanding of the arrangement, of bending electromagnets 6a, 6b, and 6c, that characterize the present invention. In FIGS. 1 through 3, the bases are omitted. The beam transport unit 15 has a beam transport duct 12, three bending electromagnets 6a, 6b, and 6c, and four quadrupole electromagnets 5a, 5b, 5c, and 5d. The beam transport unit 15 does not include a Wobbler electromagnet and a scanning electromagnet in the irradiation nozzle 8. As the reference numerals of the bending electromagnets, “6” is collectively utilized; however, in the case where the bending electromagnets are separately explained, “6a”, “6b”, and “6c” are utilized. As the reference numerals of the quadrupole electromagnets, “5” is collectively utilized; however, in the case where the quadrupole electromagnets are separately explained, “5a”, “5b”, “5c”, and “5d” are utilized. As the reference numerals of the supporting bases, “4” is collectively utilized; however, in the case where the supporting bases are separately explained, “4a” and “4b” are utilized. The bending electromagnet 6 and the quadrupole electromagnets 5 are beam transport electromagnets. The beam transport duct 12 is a vacuum duct or the like that generates a vacuum inside the apparatus. A charged particle beam 14, accelerated by a circular accelerator (synchrotron), passes through the inside of the beam transport duct 12 in which a vacuum has been generated. The bending electromagnets 6a, 6b, and 6c each bend the charged particle beam 14 toward predetermined directions. The quadrupole electromagnets 5a, 5b, 5c, and 5d each focus or defocus the charged particle beam 14 and change the beam-path-direction gradient η′ of the dispersion function η of the charged particle beam 14 so as to adjust the beam width to be within a tolerance range. The characteristic of the present invention is that the three bending electromagnets 6a, 6b, and 6c are arranged in such a way that the two bending planes of at least one pair among three pairs of the bending planes thereof are not on the same plane, not parallel with each other, and not perpendicular to each other. In other words, it suggests that when the bending planes of the three bending electromagnets 6a, 6b, and 6c are extended and crossed one another, the angle between the two bending planes of at least one pair among three pairs of the bending planes thereof is between 0 degree and 90 degrees. Moreover, it suggests that the angle between the bending magnetic fields (double-pole magnetic fields) of at least one pair among three pairs of the bending magnetic fields of the three bending electromagnets 6a, 6b, and 6c is between 0 degree and 90 degrees when the direction of the magnetic field is viewed on the xy plane that is perpendicular to the beam traveling direction. Accordingly, the respective motions of a particle, in the x direction and the y direction, that has undergone the magnetic fields of the bending electromagnet 6 and the quadrupole electromagnet 5 are not independent from each other; instead, they are coupled with each other. Here, the x axis and y axis are each perpendicular to the beam traveling direction (s direction, s axis). Because the particle beam rotational irradiation apparatus 20 rotates, the x axis and the y axis, explained herein, are defined to be one axis for the direction of the double-pole magnetic field between the magnetic poles at the inlet side of the bending electromagnet 6 at the upstreammost side, i.e., the first bending electromagnet 6a at a given rotation angle, and an another axis for the direction perpendicular to the one axis, respectively. In general, in such a particle beam rotational irradiation apparatus (rotating gantry) as disclosed in Patent Document 2 or 3, the bending electromagnets are arranged in such a way that the bending planes thereof are on the same plane, so that there is utilized the effect that the bending electromagnets and quadrupole electromagnets change the dispersion function η and the beam-path-direction gradient η′ of the dispersion function η. In a particle beam rotational irradiation apparatus (rotating gantry), by utilizing this effect, the dispersion function η is nullified or diminished to the extent that its contribution is sufficiently small so as to cancel the beam size dependence on the momentum spread at the isocenter. In this situation, the bending electromagnet can cause η; however, the quadrupole electromagnet only changes the gradient η′. The only way to keep the beam size dependence at “zero”, i.e., “η=0 and η′=0” is to change the gradient η′ by use of the quadrupole electromagnet and then cause an opposite-sign η by use of the bending electromagnet. By bending a particle beam by α degree and then bending it by −α degree on the same plane, the mode “η=0 and η′=0” can be realized only by the bending electromagnets. However, this method may cause η to become so large during the transport of a charged particle beam that the beam hits the duct. In the case of such a corkscrew-type gantry as disclosed in Patent Document 1 or 4, respective momentum spread dependences are caused in the x direction and the y direction. That is to say, the mode in which ηx≠0 and ηy≠0 takes place. However, in the case where a corkscrew-type gantry is provided with two bending electromagnets for bending a beam on a plane including the s axis and the x axis, two bending electromagnets for bending a beam on a plane including the s axis and the y axis, and a quadrupole electromagnet between the bending electromagnets, the mode in which ηx=0, ηy=0, ηx′=0, and ηy′=0 can be obtained. As a result, not only the respective momentum-spread dependences of a beam width in the x direction and the y direction can be cancelled at the isocenter, but also the respective gradients η′ (ηx′, ηy′) in the x direction and the y direction can be nullified at the isocenter. In the case where bending electromagnets are arranged in the same manner as Embodiment 1, i.e., in the case where the three bending electromagnets 6a, 6b, and 6c are arranged in such a way that the two bending planes of at least one pair among three pairs of the bending planes thereof are not on the same plane, not parallel with each other, and not perpendicular to each other, the respective motions of a particle in the x direction and the y direction are coupled with each other. Thus, the respective bending planes of the bending electromagnets 6a, 6b, and 6c do not separately nullify (or sufficiently diminish) the dispersion functions; instead, by utilizing the coupling of the respective motions of a particle in the x direction and the y direction, both the x-direction dispersion function ηx and the y-direction dispersion function ηy and both the x-direction gradient ηx′ and the y-direction gradient ηy′ of the x-direction dispersion function and the y-direction dispersion function can eventually be nullified or sufficiently be diminished at the isocenter. This method will be explained below by use of a transfer matrix utilized in beam designing calculation. When attention is paid to a single particle in a charged particle beam, the transfer matrix can be defined as in the equation (2). Characters x′ and y′ denote the respective s-direction gradients, assuming that the beam traveling direction at the particle position (x, y) is the s axis. The left-hand side of the equation (2) denotes the particle position (x, y) at the isocenter, the s-direction gradients (x′, y′) at this particle position, and the momentum spread Δp/p0. The right-hand side of the equation (2) denotes a transfer matrix M, the particle position (x, y) at the inlet of the gantry, the s-direction gradients (x′, y′) at this particle position, and the momentum spread Δp/p0. ( x x ′ y y ′ Δ p p 0 ) out = ( r 11 r 12 r 13 r 14 r 15 r 21 r 22 r 23 r 24 r 25 r 31 r 32 r 33 r 34 r 35 r 41 r 42 r 43 r 44 r 45 0 0 0 0 1 ) ( x x ′ y y ′ Δ p p 0 ) i n = M ( x x ′ y y ′ Δ p p 0 ) i n ( 2 ) The transfer matrix M is a net transfer matrix determined by the values of the magnetic fields of the quadrupole electromagnets and the bending electromagnets provided in the beam transport line (that corresponds to the beam transport unit 15 of Embodiment 1) of the rotating gantry and the drift length. Here, the drift denotes a linear section (drift section) where no magnetic field exists; the drift length denotes the length of the linear section (drift section) where no magnetic field exists. Characters r15, r25, r35, and r45 denote the x-direction dispersion function ηx, the gradient ηx′ of the dispersion function ηx, the y-direction dispersion function ηy, and the gradient ηy′ of the dispersion function ηy, respectively. Letting Mn denote the transfer matrix of elements of a magnetic field or a drift, the net transfer matrix of the rotating gantry is given by the equation (3), when there exist n drifts.M=Mn . . . M3M2M1 (3) In general, in the case of a rotating gantry in which the coupling between the respective motions of a particle in the x direction and the y direction is not utilized, the net transfer matrix of the gantry is given by the equation (4). M = ( r 11 r 12 0 0 r 15 r 21 r 22 0 0 r 25 0 0 r 33 r 34 r 35 0 0 r 43 r 44 r 45 0 0 0 0 1 ) ( 4 ) As evident from the transfer matrix in the equation (4), each of the x component or the y component of a particle position, the s-direction gradient x′ or y′ at this particle position is obtained by adding three items; therefore, the motion of the particle is equivalent to that obtained by separately calculating the respective motions in the x direction and the y direction by use of “3×3” transfer matrix. In a conventional manner, in order to separately deal with the x-direction motion and the y-direction motion, i.e., in order not to consider the coupling between the x-direction particle motion and the y-direction particle motion, r15 (the x-direction dispersion function ηx) of the transfer matrix M and r35 (the y-direction dispersion function ηy) are set to “0”, and the respective gradients r25 (gradient ηx′) and the r45 (gradient ηy′) of the s-direction dispersion functions ηx and ηx are also set to “0”. This adjustment is performed by adjusting the values of the magnetic-field intensities of a plurality of quadrupole electromagnets. Originally, a two-pole electromagnet has the objective to lead a charged particle beam to the isocenter, the bending angle thereof is preliminarily determined, and the value of the magnetic field of the two-pole electromagnet is preliminarily determined; thus, the dispersion functions ηx and ηy and their gradients ηx′ and ηy′ cannot be adjusted by changing the value of the magnetic field of the two-pole electromagnet. Unlike a two-pole electromagnet, a quadrupole electromagnet has the objective to focus or defocus a charged particle beam; however, because the value of the magnetic field of the quadrupole electromagnet is not uniquely determined, the magnetic field of the quadrupole electromagnet is an adjustment parameter for adjusting the dispersion functions ηx and ηy and their gradients ηx′ and ηy′. As described above, to date, in order to sufficiently diminish the x-direction dispersion function η(ηx) that is caused by the magnetic field of the bending electromagnet and the s-direction gradient η′(ηx′) of the dispersion function η, it has been required to provide at least one more bending electromagnet, the bending plane of which is on the same plane. Even if a quadrupole electromagnet is added, this addition alone cannot nullify both η and η′; thus, one more bending electromagnet is required. This applies to the y direction. With regard to the variation of the transfer matrix caused by the rotation of a rotating gantry, the mode of an incident beam that enter the rotating gantry is set in such a way that ηx=0, ηy=0, ηx′=0, and ηy′=0; then, if the mode in which ηx=0, ηy=0, ηx′=0, and ηy′=0 is established at the isocenter of a given rotation angle, the mode in which ηx=0, ηy=0, ηx′=0, and ηy′=0 is established at any other rotation angle. However, as described above, in a conventional beam transport line where there exists no coupling between the x-direction particle motion and the y-direction particle motion, there exists only two cases, i.e., the case where the number of the bending planes of the bending electromagnets is one and the case where the bending planes are perpendicular to each other. In contrast to a conventional beam transport line, in the arrangement condition for the bending electromagnet 6 in the beam transport unit 15 of Embodiment 1, there exists no restriction on the foregoing bending plane; therefore, the degree of freedom in the designing is high. For example, in the case where the arrangement condition for the second bending electromagnet 6b is set in such a way that the bending plane of the bending electromagnet 6b is not on the same plane as the bending plane of the first bending electromagnet 6a, not parallel with the bending plane of the first bending electromagnet 6a, or not perpendicular to the bending plane of the first bending electromagnet 6a, the magnetic field thereof is exerted in both the x direction and the y direction; thus, regardless of the rotation angle of the rotating gantry, the transfer matrix M is given by the equation (5). The transfer matrix of the second bending electromagnet 6b is expressed as M2. M 2 = ( r 11 r 12 r 13 r 14 r 15 r 21 r 22 r 23 r 24 r 25 r 31 r 32 r 33 r 34 r 35 r 41 r 42 r 43 r 44 r 45 0 0 0 0 1 ) ( 5 ) Accordingly, even when the magnetic fields of the following bending electromagnet 6 and the quadrupole electromagnets 5 are arranged under the condition that the magnetic fields are exerted in only one of the x direction and the y direction, i.e., even when as a single element, the transfer matrix is defined as in the right-hand term of the equation (4), the net transfer matrix M of the overall beam transport unit 15 undergoes the effect of the transfer matrix M2 of the second bending electromagnet 6b; therefore, each of r15, r25, r35, and r45 of the overall net transfer matrix M undergoes correlation components, i.e., other-direction components, in addition to the independent component in the x direction or the y direction. Let's take advantage of the above fact; each of r15, r25, r35, and r45 of the overall net transfer matrix M undergoes correlation components, i.e., other-direction components; thus, when the arrangement positions of, the number of, and the magnetic-field values of a plurality of quadrupole electromagnets are appropriately adjusted, the values of r15, r25, r35, and r45, i.e., the dispersion functions ηx (r15) and ηy (r35) and their gradients ηx′ (r25) and ηy′ (r45) can be nullified or sufficiently be diminished. Therefore, it is not required that in order to nullify or sufficiently diminish each of r15, r25, r35, and r45 of the overall net transfer matrix M, two bending electromagnets are arranged in such a way that the bending planes thereof are the same. In other words, unlike Patent Documents 1 and 4, it is not required that the bending electromagnet at the downstream side is disposed in such a way that the bending plane thereof is the same as the bending plane of the bending electromagnet at the upstream side. When the arrangement positions of, the number of, and the magnetic-field values of a plurality of quadrupole electromagnets are adjusted, the dispersion functions ηx and ηy and their gradients ηx′ and ηy′ can be nullified or sufficiently be diminished at the outlet of the beam transport unit 15 and at the isocenter, even in the case of such a particle beam rotational irradiation apparatus in which the beam path of the beam transport unit 15 is twisted, i.e., even in the case of the particle beam rotational irradiation apparatus 20 as described in Embodiment 1; therefore, it is made possible to diminish the momentum-spread dependency of a beam width, i.e., the momentum spread Δp/p0 can be cancelled to the extent that its contribution to the beam width is sufficiently small. Mathematically, the desired solution for nullifying or sufficiently diminishing the four components r15, r25, r35, and r45 in the elements of the matrix exists at a high probability. It goes without saying that when the number of provided quadrupole electromagnets is more than 4, the adjustment is facilitated. Moreover, the particle beam rotational irradiation apparatus can be realized with three bending electromagnets, while four bending electromagnets are required for a corkscrew-type. Still moreover, in the case where as illustrated in FIG. 4, the particle beam rotational irradiation apparatus 20 is installed in such a way that the front end of the irradiation nozzle 8 protrudes toward the irradiation room, it is made possible to realize a small-size and open-type rotating gantry. FIG. 4 is a perspective view of the particle beam rotational irradiation apparatus in FIG. 1, taken from the irradiation room. A rotation front cover 16 is a cover mounted on the rotating unit 1 in such a way as to cover the front side of the rotating unit 1 and rotates as the rotating unit 1 rotates. The rotation front cover 16 is disposed in a rotatable manner in the through-hole opened in an irradiation room wall 17. The particle beam rotational irradiation apparatus 20 according to Embodiment 1 is an open-type rotating gantry and is characterized in that the last bending electromagnet 6c is disposed obliquely with respect to the irradiation room wall 17. The particle beam rotational irradiation apparatus 20 according to Embodiment 1 is an open-type rotating gantry in which a treatment table 9 can be disposed on an irradiation room base 18; therefore, unlike such a particle beam rotational irradiation apparatus in which the treatment table 9 is disposed in the narrow inner chamber of the rotating gantry, the particle beam rotational irradiation apparatus 20 can facilitate patient positioning work and hence the throughput of the particle beam therapy is raised. In Embodiment 1, as can be seen from FIG. 3, the beam transport unit 15 is configured in such a way that the respective bending planes of the bending electromagnets 6a, 6b, and 6c are not on the same plane, not parallel with one another, and not perpendicular to one another. The respective magnetic fields of the quadrupole electromagnets 5a, 5b, 5c, and 5d of the beam transport unit 15 are set in such a way that the beam width (beam size) of the charged particle beam 14 at the outlet of the beam transport unit 15, i.e., at the outlet of the downstreammost bending electromagnet 6c falls within the tolerance range, i.e., in such a way that the dispersion functions ηx and ηy of the charged particle beam 14 and their gradients ηx′ and ηy′ are nullified or sufficiently diminished at the outlet of the bending electromagnet 6c. In the particle beam rotational irradiation apparatus 20 according to Embodiment 1, the beam transport unit 15 is configured in such a way that the respective bending planes of the bending electromagnets 6a, 6b, and 6c are not on the same plane, not parallel with one another, and not perpendicular to one another, and the respective magnetic fields of the quadrupole electromagnets 5a, 5b, 5c, and 5d of the beam transport unit 15 are set in such a way that the dispersion functions ηx and ηy of the charged particle beam 14 and their gradients ηx′ and ηy′ are nullified or sufficiently diminished at the outlet of the downstreammost bending electromagnet 6c; therefore, the small-size and open-type particle beam rotational irradiation apparatus 20 can be realized. Not only in the case where the beam transport unit 15 is configured in such a way that the respective bending planes of the bending electromagnets 6a, 6b, and 6c are not on the same plane, not parallel with one another, and not perpendicular to one another, but also in the case where the three bending electromagnets 6a, 6b, and 6c are arranged in such a way that the bending planes of at least one pair among three pairs of the bending planes thereof are not on the same plane, not parallel with each other, and not perpendicular to each other, the foregoing explanation is established, because regardless of how to set the x axis and the y axis, the magnetic fields are exerted in such a way that the motions of a particle in the x direction and the y direction are coupled with each other, i.e., any one of the transfer matrix Mn is given by the equation (5). Therefore, in this case, as well, the respective magnetic fields of the quadrupole electromagnets 5a, 5b, 5c, and 5d of the beam transport unit 15 are set in such a way that the dispersion functions ηx and ηy of the charged particle beam 14 and their gradients ηx′ and ηy′ are nullified or sufficiently diminished at the outlet of the beam transport unit 15, i.e., at the outlet of the downstreammost bending electromagnet 6c; thus, the small-size and open-type particle beam rotational irradiation apparatus 20 can be realized. However, as a matter of course, the relationship in which the respective bending planes of the bending electromagnets 6a, 6b, and 6c are not on the same plane, not parallel with one another, or not perpendicular to one another raises, compared with the other manner, the degree of freedom in adjusting the magnetic fields of the quadrupole electromagnets 5a, 5b, 5c, and 5d and hence facilitates obtaining desired beam parameters (a beam size and a gradient with respect to the beam traveling direction (the beam center axis)). Moreover, by utilizing three bending electromagnets, the number of the bending electromagnets can be minimized compared with the case where four bending electromagnets are utilized; thus, the low-cost particle beam rotational irradiation apparatus 20 can be realized. Still moreover, as the bending electromagnet 6, a bending electromagnet in which four-pole components including the magnetic field at the end are utilized is adopted so as to replace part of the quadrupole electromagnets 5 and to decrease the number of the quadrupole electromagnets 5, so that the cost can be reduced. As described above, the particle beam rotational irradiation apparatus 20 according to Embodiment 1 is provided with the irradiation nozzle 8 that irradiates the charged particle beam 14, the beam transport unit 15 that transports the charged particle beam 14 to the irradiation nozzle 8, and the rotating unit 1 that can rotate around the isocenter; the beam transport unit 15 has three or more bending electromagnets 6. The bending electromagnets 6a, 6b, and 6c are arranged in such a way that in the case where as a pair of bending planes, any two of the bending planes of the bending electromagnets 6 are selected, the two bending planes of at least one pair of bending planes are not on the same plane, not parallel with each other, and not perpendicular to each other; therefore, the respective motions of a charged particle in the x direction and the y direction are coupled with each other; thus, because it is not required to perform the adjustment both in the x direction and in the y direction, the working space for performing patient positioning work can be widened even in the case of a small-size particle beam rotational irradiation apparatus and hence the throughput of particle beam therapy can be improved. FIG. 5 is a side view illustrating a particle beam rotational irradiation apparatus according to Embodiment 2 of the present invention; FIG. 6 is an elevation view illustrating a particle beam rotational irradiation apparatus according to Embodiment 2 of the present invention. A particle beam irradiation apparatus 20 according to Embodiment 2 has a structure that is not an open type. Because the last two bending electromagnets 6b and 6c are not arranged in a row, the width of an inner chamber, which is a tunnel portion in which a patient who undergoes a therapy is situated, can be enlarged in comparison with a particle beam rotational irradiation apparatus according to Patent Document 1 or 4; thus, a relatively open rotating gantry can be realized. Because being not an open type, the particle beam rotational irradiation apparatus 20 according to Embodiment 2 has an advantage that in comparison to the particle beam rotational irradiation apparatus 20 according to Embodiment 1, the irradiation nozzle 8 can readily be supported. Because in the particle beam rotational irradiation apparatus 20 according to Embodiment 2, the width of an inner chamber, which is a tunnel portion in which a patient who undergoes a therapy is situated, can be enlarged in comparison with a particle beam rotational irradiation apparatus according to Patent Document 1 or 4, a relatively open rotating gantry can be realized; thus, there can be realized a particle beam rotational irradiation apparatus that can raise the throughput of particle beam therapy and is downsized. Even though being a small-size particle beam rotational irradiation apparatus, the particle beam rotational irradiation apparatus 20 according to Embodiment 2 makes it possible to widen the working space where work in the vicinity of the irradiation nozzle 8 is performed and hence the throughput of particle beam therapy can be improved. Moreover, a monitoring apparatus provided utilizing the wide space facilitates irradiation while performing moving body tracking. In Embodiment 1, it has been explained that when the arrangement positions of, the number of, and the magnetic-field values of a plurality of quadrupole electromagnets are adjusted, the dispersion functions ηx and ηy and their gradients ηx′ and ηy′ can be nullified or sufficiently be diminished at the outlet of the beam transport unit 15 and at the isocenter, even in the case of a particle beam rotational irradiation apparatus in which the beam path of the beam transport unit 15 is twisted, i.e., even in the case of the particle beam rotational irradiation apparatus 20 as described in Embodiment 1, and hence the momentum-spread dependence of a beam width can be diminished. In Embodiment 3, it will be explained that by contriving the arrangement positions of the magnetic poles of the quadrupole electromagnets 5, it is made possible to transport a beam having no xy correlation at the isocenter. At first, the relationship between the magnetic pole 21 of the quadrupole electromagnet 5 and the focusing or defocusing direction of the charged particle beam 14 will be explained. FIG. 7 is a diagram for explaining the acting direction of a quadrupole electromagnet of the present invention. The quadrupole electromagnet 5 is an electromagnet having four magnetic poles 21a, 21b, 21c, and 21d and coils 22a, 22b, 22c, and 22d wound around those magnetic poles; these magnetic poles and coils focus or defocus the charged particle beam 14 in the broken-line arrow 23 or 24. The direction from the center to the outside is a defocusing direction, and the direction from the outside to the center is a focusing direction. The directions of the broken-line arrows 23 and 24 will be referred to as acting directions 23 and 24 in each of which focusing and defocusing occurs. In general, in a conventional beam transport line, the quadrupole electromagnet 5 is disposed in such a way that the acting directions 23 and 24 are each perpendicular to or parallel with the bending plane of the bending electromagnet 6 and is utilized in such a way that the x-direction motion and the y-direction motion of a particle are not coupled with each other. In other words, the quadrupole electromagnet 5 in a conventional beam transport line is disposed in such a way that as illustrated in FIG. 9, the bending plane 55 of the bending electromagnet 6 is parallel with the acting direction 23 and perpendicular to the acting direction 24. FIG. 9 is a diagram illustrating the arrangement of the acting directions of the magnetic poles in a conventional quadrupole electromagnet. Next, there will be explained how to deal with a beam in a rotating gantry. The post-drift positions and the gradients of the traveling directions of respective charged particles that form a beam are determined by the equation (2); however, the parameters (the beam size, the beam gradient with respect to the beam traveling direction (beam center-axis direction) of a beam, which is an aggregate of charged particles, can be described by a σ matrix, which is a statistic quantity of the particle distribution, and is given by the equation (6). σ = ( σ 11 σ 12 σ 13 σ 14 σ 21 σ 22 σ 23 σ 24 σ 31 σ 32 σ 33 σ 34 σ 41 σ 42 σ 43 σ 44 ) = ( 〈 x 2 〉 〈 xx ′ 〉 〈 xy 〉 〈 xy ′ 〉 〈 xx ′ 〉 〈 x ′ 2 〉 〈 x ′ y 〉 〈 x ′ y ′ 〉 〈 xy 〉 〈 x ′ y 〉 〈 y 2 〉 〈 yy ′ 〉 〈 xy ′ 〉 〈 x ′ y ′ 〉 〈 yy ′ 〉 〈 y ′ 2 〉 ) ( 6 ) The relationship between the σ matrix and the transfer matrix M is given by the equation (7). The equation (7) is shown in a document (Jean Buon LAL/RT96-03 April 1996). In the case of rotating gantries disclosed in Patent Documents 1 through 4, by rotating the coordinate system along with the rotation of the rotating gantry, there can be designed the transport of a beam in which, ideally, there are no xy correlation components of the beam parameters (the beam size, the beam gradient with respect to the beam traveling direction (beam center-axis direction).σ2=Mσ1MT (7) Next, there will be explained the arrangement of the positions of the magnetic poles in the foregoing quadrupole electromagnet 5. FIG. 8 is a diagram illustrating the arrangement of the acting directions of the magnetic poles in a quadrupole electromagnet according to Embodiment 3 of the present invention; FIG. 10 is a diagram illustrating the mounting position of the quadrupole electromagnet in FIG. 8. In the rotating gantry (particle beam rotational irradiation apparatus 20) according to the present invention, even when as the x axis and the y axis of the charged particle beam 14, whatever two axes that are perpendicular to each other are selected, the components (r13, r14, r23, r24, r31, r32, r41, and r42), in the transfer matrix M, that each generate the correlations between the x-direction particle motion and the y-direction particle motion do not become “0”; therefore, even when there exists no xy correlation in σ1 (the spread of a beam launched into the gantry), the equation (7) suggests that the components (σ13=σ31, σ14=σ41, σ23=σ32, and σ24=σ42), in the beam parameters (the beam size, the beam gradient with respect to the beam traveling direction (the direction of the beam-center axis)) of σ2 (the spread of a beam that exits from the gantry), that each indicate the xy correlations are not “0”. In this case, when both the x-direction beam size (beam width) and the y-direction beam size of the charged particle beam 14 are reduced, especially in the section between the last bending electromagnet (the bending electromagnet 6c in FIG. 10) and the isocenter, one of the beam sizes or its gradient always provides an effect to the other direction; thus, the adjustment becomes difficult in both directions. Moreover, it is required to make a treatment plan for which a beam having the xy correlation is taken into consideration. That is to say, in the particle beam rotational irradiation apparatus 20 according to Embodiment 1, the foregoing problem is still remaining. The foregoing problem does not exist in a conventional rotating gantry where as the x axis and the y axis of the charged particle beam 14, two axes, i.e., an axis that is horizontal to the bending plane of the bending electromagnet 6 and an axis that is vertical to the bending plane of the bending electromagnet 6 are selected, the σ matrix can be considered as a matrix where there exists no coupling, i.e., a matrix, the x direction and the y direction components of which are independent from each other in such a way that σ13=σ31=0. σ14=σ41=0, σ23=σ32=0, and σ24=σ42=0. Accordingly, as illustrated in FIG. 8, the quadrupole electromagnet 5 is disposed with its magnetic-pole positions arranged in such a way that the focusing direction and the defocusing direction thereof are neither parallel with nor perpendicular to the bending planes 25 and 26 of the bending electromagnets 6 situated at both sides thereof (at the immediate upstream side and the immediate downstream side of the quadrupole electromagnet 5). In the case where the quadrupole electromagnet 5 is disposed in this manner, the components that generate the particle-motion xy correlation in the transfer matrix M for the quadrupole electromagnet 5 are not “0” and hence act on the components, caused by the bending electromagnet disposed under the condition of the present invention, that generate the particle-motion xy correlation; therefore, the components that generate the particle-motion xy correlation can be reduced at the last bending electromagnet (the bending electromagnet 6c in FIG. 10) to the extent that the components (σ13=σ31, σ14=σ41, σ23=σ32, and σ24=σ42) is “0” or the effect on the beam size can sufficiently be neglected. FIG. 8 illustrates an example in which the angle between the bending plane 25 and the acting direction 23 is θ1 (0<θ1<90°) and the angle between the bending plane 26 and the acting direction 23 is θ2 (0<θ2<90°). Moreover, in comparison with a conventional particle beam rotational irradiation apparatus in which only the magnetic-field intensity of the quadrupole electromagnet 5 and the arrangement position, on the beam path, of the quadrupole electromagnet 5 are dealt with as designing parameters, the particle beam rotational irradiation apparatus 20 according to Embodiment 3 makes it possible to perform the designing by use of the rotation angle (the angles θ1 and θ2 in FIG. 8) of the quadrupole electromagnet 5 with respect to the beam center, as adjustment parameters; therefore, the degree of freedom in the designing can be raised. Conventionally, the rotation angle of the quadrupole electromagnet 5 is 90° (θ3), as illustrated in FIG. 9, and does not work as an adjustment parameter. In the case where as illustrated in FIG. 11, a rotation-driving mechanism 30 capable of driving the quadrupole electromagnet 5 in a rotating manner is provided, the rotation angle θ of the quadrupole electromagnet 5 can be adjusted even when due to a magnetic-field error or the like, the beam size differs from the designed beam size; thus, it can be expected that the degree of freedom in the adjustment of the beam size is raised. FIG. 11 is a diagram illustrating the rotation-driving mechanism of a quadrupole electromagnet according to Embodiment 3 of the present invention. The rotation-driving mechanism 30 is provided with a ring 31, a supporting member 32 that connects the quadrupole electromagnet 5 with the ring 31, and a rotation-driving device 33 that makes the ring 31 rotate. In general, in the beam transport line, various adjustment methods are utilized for the purpose of correcting the fact that due to the effect of a magnetic-field error, the beam orbit center differs from the designed beam orbit center. For example, the orbit correction (a defocused orbit center is corrected to be the designed orbit center) through a method in which a steering electromagnet (two-pole electromagnet) is disposed in the beam transport line and is energized, a method in which an auxiliary coil is wound around the bending electromagnet 6 in the beam transport line and is energized by a current, a method in which as an offset, a current corresponding to a magnet field required for the orbit correction is made to flow in the bending electromagnet 6 of the beam transport line, and the like. In general, when the x-direction and y-direction beam parameters (the beam size, the beam gradient with respect to the beam traveling direction (the direction of the beam-center axis)) are not coupled with each other, the orbit correction is performed in both the x axis and the y axis; in order to implement this, a magnet for exerting a magnetic field in the x direction and a magnet for exerting a magnetic field in the y direction are required (in some cases, a single magnet, which generates two-pole magnetic field in the both directions, is utilized). In some cases, the foregoing method is utilized also in the particle beam rotational irradiation apparatus 20 according to the present invention; however, in that case, even when the foregoing correction method is utilized at the place where there exists the xy correlation, the xy correlation provides an effect to not only the y-direction beam but also the x-direction beam when a magnetic field is exerted in the x direction; therefore, in some cases, the adjustment time becomes long. In particular, because the charged particle beam 14 launched from the irradiation nozzle 8 is a beam that is immediately utilized in the therapy, it is desired that the xy correlation in the charged particle beam 14 is sufficiently diminished and then the orbit thereof is corrected. Accordingly, after the correlation between the axis that is vertical to the bending plane of the last bending electromagnet (the bending electromagnet 6c in FIG. 10) and the axis that is horizontal to the bending plane of the last bending electromagnet is cancelled by the second-from-the-last bending electromagnet (the bending electromagnet 6b in FIG. 10), a steering electromagnet is disposed in a setting section (between the second-from-the-upstream-side bending electromagnet 6b and the third-from-the-upstream-side bending electromagnet 6c) so that the orbit correction can be performed separately in the both axes; as a result, the orbit of the charged particle beam 14 can readily be corrected. In this case, the quadrupole electromagnets 5 having the magnetic-pole arrangement illustrated in FIG. 8 are disposed before the second-from-the-last bending electromagnet (the bending electromagnet 6b in FIG. 10); for example, in the case where as illustrated in FIG. 10, the gantry has three bending electromagnets 6, the quadrupole electromagnets 5 are disposed in a setting section 34 between the first-from-the-upstream-side bending electromagnet 6a and the second-from-the-upstream-side bending electromagnet 6b, in a setting section 35 before the first-from-the-upstream-side bending electromagnet 6a, or in both the setting section 34 and the setting section 35. The quadrupole electromagnet 5 having the magnetic-pole arrangement illustrated in FIG. 8 also rotates along with the rotating gantry, while maintaining the relative position between the rotating gantry and itself. In the particle beam rotational irradiation apparatus 20 according to Embodiment 3, at least one quadrupole electromagnet 5 is disposed in the beam transport unit 15 in such a way that the focusing direction and the defocusing direction thereof are neither parallel with nor perpendicular to the respective bending planes 25 and 26 of the bending electromagnets 6 situated at both sides thereof (at the immediate upstream side and the immediate downstream side thereof); therefore, it is made possible to transport the charged particle beam 14 having no xy correlation to the isocenter. Moreover, in the particle beam rotational irradiation apparatus 20 according to Embodiment 3, the quadrupole electromagnets 5 is disposed in the drift section (the setting section 34 or 35) before the second-from-the-last bending electromagnet (the bending electromagnet 6b in FIG. 10) in the beam transport unit 15 in such a way that the focusing direction and the defocusing direction thereof are neither parallel with nor perpendicular to the respective bending planes 25 and 26 of the bending electromagnets 6 situated at both sides thereof (at the immediate upstream side and the immediate downstream side thereof); therefore, it is made possible to transport the charged particle beam 14 having no xy correlation to the isocenter and to readily correct the orbit of the beam. Next, the installation of an irradiation-field-enlargement electromagnet or a scanning electromagnet will be described. In the case where the last bending electromagnet 6c cancels the xy correlation, a pair of irradiation-field-enlargement electromagnets (e.g., Wobbler electromagnets) for enlarging a beam or a pair of scanning electromagnet for scanning a beam is disposed in a section from the last bending electromagnet 6c to the isocenter, i.e., in the irradiation nozzle 8. In the case where in this situation, the x axis and the y axis of the charged particle bean 14 are an axis that is perpendicular to the bending plane of the last bending electromagnet 6c and an axis that is parallel with the bending plane of the last bending electromagnet 6c, the condition in which there exists no xy correlation is maintained during the therapy, when the foregoing irradiation-field-enlargement electromagnet or scanning electromagnet is disposed in such a way that the bending plane thereof is perpendicular to or parallel with the bending plane of the last bending electromagnet 6c; therefore, it is not required to make a treatment plan for which the xy correlation is taken into consideration, and hence it can be expected that the time for making the treatment plan is shortened. In the particle beam rotational irradiation apparatus 20 according to Embodiment 3, the irradiation-field-enlargement electromagnet or the scanning electromagnet, which is disposed in a section from the last bending electromagnet 6c in the beam transport unit 15 to the isocenter, is provided in such a way that the direction of the two-pole magnetic field thereof is parallel with or perpendicular to the direction of the magnetic field of the last bending electromagnet 6c; therefore, it is made possible to transport the charged particle beam 14 having no xy correlation to the isocenter and to readily adjust the beam size at the isocenter and the beam center position. Furthermore, in the case where the correlation between the x-direction particle motion and the y-direction particle motion can sufficiently be diminished only by disposing at least one quadrupole electromagnet which is made in such a way that the focusing direction and the defocusing direction thereof are neither parallel with nor perpendicular to the respective bending planes of the bending electromagnets situated at both sides thereof in a drift section before the second-from-the-last bending electromagnet (the bending electromagnet 6b in FIG. 10), the xy correlation of the beam in the setting section 36 has already been cancelled (in the direction of the magnetic field of the last bending electromagnet 6c); therefore, it is made possible to dispose an irradiation-field-enlargement electromagnet or a scanning electromagnet in the setting section 36. As a result, the irradiation nozzle 8 is downsized, and hence it can be expected that the rotation diameter of the gantry is diminished. The orbit correction is performed with a two-pole electromagnet; therefore, a pair of beam scanning electromagnets (scanning electromagnets) or irradiation-field-enlargement electromagnets (Wobbler electromagnets) that scans a beam in two vertical-axis directions is disposed in a section between the second-from-the-last bending electromagnet 6b and the last bending electromagnet 6c and then an auxiliary coil is wound around the magnet of at least one of the beam scanning electromagnet or the irradiation-field-enlargement electromagnet or offset current energization is performed, so that the orbit correction in which no xy coupling occurs can also be performed. In this situation, the beam scanning electromagnet or the irradiation-field-enlargement electromagnet is disposed in such a way that the direction of the two-pole magnetic field is horizontal or vertical to the bending plane of the last bending electromagnet 6c. In this case, the arrangement from the irradiation nozzle 8 is changed, and the beam scanning electromagnet or the irradiation-field-enlargement electromagnet is integrated in the beam transport unit 15. In another particle beam rotational irradiation apparatus 20 according to Embodiment 3, the beam scanning electromagnet or the irradiation-field-enlargement electromagnet is disposed in a section between the last bending electromagnet 6c in the beam transport unit 15 and the second-from-the last bending electromagnet 6b is provided in such a way that the direction of the two-pole magnetic field thereof is parallel with or perpendicular to the direction of the magnetic field of the last bending electromagnet 6c; therefore, it is made possible to transport the charged particle beam 14 having no xy correlation to the isocenter and to readily correct orbit-center displacement caused by a magnetic-field error. Moreover, another particle beam rotational irradiation apparatus 20 according to Embodiment 3 is configures in such a manner as described above; therefore, because the beam scanning electromagnet or the irradiation-field-enlargement electromagnet can have a function as a steering magnet, the number of the steering magnets can be reduced. Various modifications and alterations of this invention will be apparent to those skilled in the art without departing from the scope and spirit of this invention, and it should be understood that this is not limited to the illustrative embodiments set forth herein. |
|
claims | 1. A core catcher for use in a boiling water nuclear plant which has:a base mat;a reactor building built on a part of the base mat;a containment vessel provided in the reactor building, built on the base mat and having a total height of not exceeding 29.5 m to a lower end of a top slab;a core;a reactor pressure vessel holding the core;a dry well constituting a part of the containment vessel and holding the reactor pressure vessel;a pedestal connected to the base mat and supporting the reactor pressure vessel through a vessel skirt and a vessel support;a wet well constituting a part of the containment vessel, the wet well being provided around the pedestal, holding a suppression pool in a lower part thereof, and having a wet well gas phase at an upper part thereof;LOCA vent pipes provided in a sidewall of the pedestal and connecting the dry well to the suppression pool;a lower dry well which is a space in the dry well, is located below the vessel skirt and the reactor pressure vessel and is surrounded by the sidewall of the hollow cylindrical pedestal and the part of the base mat, which lies inside the sidewall of the pedestal;control rod drives provided in the lower dry well and connected to a lower part of the reactor pressure vessel; anda control rod drive handling equipment provided in the lower dry well and below the control rod drives;the core catcher comprising:a main body including:a distributor arranged on the part of the base mat in the lower dry well, a basin arranged on the distributor,cooling channels arranged on a lower surface of the basin, having inlets connected to the distributor and extending in radial directions, anda riser connected to outlets of the cooling channels and extending upward in vertical direction;a lid connected to an upper end of the riser and covering the main body;a cooling water injection pipe connected at an inlet end to the suppression pool, penetrating the sidewall of the pedestal, connected at an outlet end to the distributor, and configured to supply suppression pool water to the distributor; andchimney pipes connected at an inlet end to the riser, penetrating the sidewall of the pedestal, and having an outlet end located above the upper end of the riser and submerged in the suppression pool water at a level lower than a minimum water level at a time of an accident,wherein the upper ends of the main body and the lid are at heights lower than lower end of the control rod drive handling equipment, as measured from upper end of the base mat,wherein the cooling water injection pipe, the distributor, the cooling channels, the riser and the chimney pipes are kept communicated with the suppression pool at all times, and always filled with the suppression pool water, the core catcher further comprising:a sump; anda sump riser extending upward along a sidewall of the sump, whereinthe basin and parts of the cooling channels surround the peripheries of the sump in conformity with a shape of the sump. 2. The core catcher according to claim 1, whereina refractory layer is provided along an upper surface of the basin and along a side of the riser. 3. The core catcher according to claim 2, wherein a sacrificial layer is provided along the refractory layer. 4. The core catcher according to claim 1, wherein the upper ends of the main body and the lid are positioned lower than a level 1.7 m above the upper end of the base mat. 5. The core catcher according to claim 4, whereina refractory layer is provided along an upper surface of the basin and along a side of the riser. 6. The core catcher according to claim 5, wherein a sacrificial layer is provided along the refractory layer. |
|
abstract | A fine motion control rod drive mechanism handling apparatus attaches a fine motion control rod drive mechanism having a control rod drive mechanism body, a spool piece, a motor bracket and a motor unit to a reactor pressure vessel, and detaches it from the reactor pressure vessel. The fine motion control rod drive mechanism handling apparatus is provided with a bolt wrench assembly, a motor unit attachment mounted to the bolt wrench assembly, for meshing a first gear of a first gear coupling on a spool piece side with a second gear of a second gear coupling on a motor unit side, and a rotation mechanism mounted to the bolt wrench assembly, for rotating the motor unit attachment. |
|
abstract | A bath containing nickel ions and formic acid is injected into a film-forming aqueous solution flowing in a circulation pipe connected to a feed water pipe made of carbon steel in a BWR plant. This solution is supplied into the pipe through the circulation pipe, and a nickel metal film is formed on an inner surface of the pipe. After the film is formed, a film-forming aqueous solution containing iron (II) ions, formic acid, nickel ions, hydrogen peroxide, and hydrazine is supplied to the pipe. A nickel ferrite film is formed on the surface of the nickel metal film in the pipe. The nickel ferrite film comes into contact with water containing dissolved oxygen at or above 150° C. to transform the nickel metal film into a nickel ferrite film. A thick nickel ferrite film is formed on the inner surface of the feed water pipe. |
|
046577319 | abstract | Water containing cesium, such as the reactor coolant of a pressurized water nuclear reactor, is purified by passing the water through an ion exchange resin bed containing a highly cross-linked, macroporous, strong acid resin derived from a monovinylidene aromatic and a cross-linking monomer compolymerizable therewith. The cross-linking monomer is employed in amounts of at least 12 mole percent based on the total weight of the monovinylidene aromatic and cross-linking monomer. |
abstract | A method of compensating mechanical, magnetic and/or electrostatic inaccuracies in a scanning charged particle beam device is described. The method includes an alignment procedure, wherein the following steps are conducted: compensating 4-fold astigmatism with an element having at least 8-pole compensation capability, wherein the aligning and compensating steps of the alignment procedure act on a charged particle beam with beam dimensions in two orthogonal directions each of at least 50 μm and coaxially aligned with at least the element having at least the 8-pole compensation capability. |
|
047028785 | summary | BACKGROUND OF THE INVENTION 1. Field of the Invention This invention relates to an apparatus for servicing a steam generator and, more particularly, to an apparatus for remotely inspecting and cleaning areas between the tubes on a tube sheet in a nuclear steam generator. 2. Description of the Prior Art In a pressurized water nuclear powered electric generating system, the heat generated by the nuclear reactor is absorbed by a primary coolant that circulates through the reactor core and is utilized to generate steam in a steam generator. The steam generator typically is an upright cylindrical pressure vessel with hemispherical end sections. Such a generator typically comprises an outer vertically oriented shell, a horizontal plate called a tube sheet adjacent the lower end of the shell, a bundle of vertical U-shaped tubes supported by the tube sheet, and a wrapper barrel inside of the outer shell surrounding the tubes and extending from the upper portion of the shell downwardly to a predetermined point above the tube sheet. The wrapper barrel forms a narrow annulus inside the shell and generally extends down to a point six inches above the tube sheet. The outer cylindrical shell is provided with openings of limited size called handholes which are located 18 inches above the tube sheet. These handhole openings are covered during operation of the steam generator but may be opened when the generator is shut down to permit access to the area inside for maintenance purposes. Occasionally, during maintenance inside the steam generator, objects such as bolts are dropped and have to be removed. Due to the limited space within the generator, the annulus between barrel wrapper and the shell generally is only two inches wide and the space beneath the bottom of the barrel wrapper is only six inches high. The bundle of tubes supported by the tube sheet also are very closely spaced together. Therefore, it is difficult to insert a device into the generator and to pick up objects between the pipes on the tube sheet. While certain inspection and maintenance devices for steam generators are known, none of these devices are adapted to be able to search between the tubes of a steam generator and also retrieve loose objects in the narrow spaces between the tubes. For example, Creek et al. U.S. Pat. No. 4,515,747 discloses a remotely operated maintenance and inspection device which includes a wheeled transporter which is inserted through a handhole opening in the shell of a steam generator and positioned on the tube sheet between the tube bundle and the inner wall of the shell. A cable is then attached to each end of the transporter and extends around the bundle of tubes and out through the handhole opening whereby suitable power means may be utilized to move the transporter around the periphery of the tube bundle. A TV camera may be mounted on the transporter to inspect the tubes of the bundle. A sludge lance also may be mounted upon the transporter to break up and liquify deposits at the base of the tubes of the bundle. The Creek et al. device, however, does not provide any means for employing a device which can be manipulated through the bundle of tubes to pick up objects on the tube sheet. Moreover, since the transporter and its associated cable contact the tubes of the tube bundle, there is a risk that the tubes might be damaged during operation of the device. Another type of prior art steam generator inspection device is disclosed in McShane et al. U.S. Pat. No. 4,355,536. This device basically comprises a probe mounted on the end of a ribbon-like probe carrier which is fed into the shell of a steam generator through the handhole opening and deflected by a carriage placed in alignment with a lane of tubes whereby the probe may measure the accumulation of sludge adjacent the bottom of the tubes. No means is provided on the device, however, for picking up and retrieving objects from the tube sheet. Accordingly, a need still exists for a device capable of inspecting or searching the tube sheet in the areas between the vertical tubes and which also is adapted to pick up any loose objects between the tubes on the tube sheet. SUMMARY OF THE INVENTION Accordingly, it is an object of the present invention to provide a device for searching and retrieving objects on a tube sheet of a steam generator in the narrow spaces between the tubes and which readily may be inserted through the handhole opening of the shell of the steam generator and placed in operation on the tube sheet. It is another object of the present invention to provide a device for searching and retrieving objects between the tubes on a tube sheet in a steam generator which is small enough in size to be placed in operating position inside the shell of the generator without dismantling any parts of the generator. It is still another object of the present invention to provide a device for searching and retrieving objects on the tube sheet between the tubes of a steam generator without risk of damage to the tubes. It is a further object of the present invention to provide a device for searching and retrieving objects on the tube sheet between the tubes of a steam generator which is of relatively simple construction and reliable in operation. To achieve the foregoing and other objects of the present invention, and in accordance with the purpose of the invention, there is provided a device for searching and retrieving objects in a steam generator such as a nuclear steam generator wherein the generator includes an outer cylindrical shell, a horizontal tube sheet adjacent the lower end of the shell, a bundle of vertical tubes supported by the tube sheet, a wrapper barrel surrounding the tubes extending from the upper portion of the shell downwardly to a predetermined point above the tube sheet to form an annulus inside the shell and an opening called a handhole through the shell to the annulus. The search and retrieval device includes a sled adapted to be passed through the shell opening and down through the annulus to the surface of the tube sheet. A flexible tube is connected at one end to the sled and has its other end extending out through the shell opening. The tube is of sufficient length to be freely movable into and out of the shell opening to move the sled along the periphery of the tube sheet. A probe such as a TV camera or a fiberscope and a gripper are mounted on the sled for searching and retrieving objects. Adjusting means including a rotatable turret are mounted on the sled for changing the operating positions of the probe and the gripper so that they may be operated between the tubes on the tube sheet. Control means extend from outside the shell through the flexible tube for operating the probe, the gripper and the adjusting means. These, together with other objects and advantages, which subsequently will be apparent, reside in the details of construction and operation of the invention as more fully hereinafter described and claimed, reference being made to the accompanying drawings forming a part hereof wherein like numerals refer to like parts throughout. |
summary | ||
042882923 | summary | CROSS-REFERENCE TO RELATED APPLICATION Reference is made to U.S. patent application filed concurrently herewith in the name of R. T. Hardin, entitled "Connecting Apparatus For Limited Rotary Or Rectilinear Motion", assigned to the Westinghouse Electric Corporation, and hereby incorporated by reference. That application discloses apparatus providing an alternative to that disclosed herein, particularly where lateral spatial limitations are not as severe. 1. Field of the Invention This invention relates to apparatus for providing connection between two members having relative motion in a horizontal plane, especially members having limited lateral space therebetween, particularly adaptable to the rotating plugs of the reactor vessel head of a nuclear reactor. 2. Description of the Prior Art Certain types of nuclear reactors require that refueling and maintenance operations be performed remotely, that is, without removal of the reactor vessel head. A plurality of rotating plugs have been provided for these operations which, through controlled rotation, allow remote access to core locations. It is important, however, that electrical and other connections be maintained to the various plugs and reactor components during relative motion. This requirement is complicated by the large number of connections and further by spatial limitations in the reactor head vicinity. This is particularly a concern with respect to the small rotating plug through which an in-vessel transfer machine must operate to shuffle fuel between the core and storage locations without removal of the reactor vessel head. Further, it is desirable that electrical signal level separation be provided among the various cables. In the prior art, reactor designers faced with these limitations have utilized apparatus similar to a split tank-type track having a fixed end and a movable end. The cables are contained within supporting plates or carriers, having openings bored to the size of the cable diameters, which are positioned between a plurality of chain links. The links are assembled to a close tolerance, allowing bending of the apparatus as necessary during motion. Although such apparatus properly perform the cable connection function, they are not without deficiencies. The tracks require a significant amount of space, which can affect the design of the connected components or limit the number of cables utilized. Further, the track system requires a large number of individual interlocking components. The large number of pins and sliding joints adds a potential for failure and a significant cost to the reactor systems. In other applications where it is desirable to provide electrical connections between movable components, systems have been proposed which have similar deficiencies, particularly in the complexity and relatively large number of components required. Two such applications are described in U.S. Pat. No. 3,351,612 in the name of R. L. Guntner, and U.S. Pat. No. 3,537,059 in the name of E. W. Howe. It therefore is desirable to provide apparatus which allows continuous connection of the large number of cables or hoses during rotation and refueling operations which also minimizes interference with personnel and other equipment while additionally providing electrical signal separation. SUMMARY OF THE INVENTION This invention discloses apparatus for providing connection between two members movable with respect to each other in a horizontal plane in a limited rectilinear or rotary fashion. The apparatus is particularly adaptable to the small rotating plug of a nuclear reactor, and offers high reliability as a result of the relatively few moving components. In a preferred embodiment the invention includes a number of upper vertically oriented plates or surfaces, laterally spaced from one another, each having a laterally extending upper restraint. It further includes a second set of lower vertical plates or surfaces, also laterally spaced from one another, and each having a laterally extending lower restraint. Each of these sets of vertical surfaces is affixed to one of the movable members, so that the vertical surfaces or plates are aligned at common lateral positions. The aligned sets are also vertically spaced from one another so as to provide a gap therebetween. A number of flexible connecting conduits, or cables or hoses, are positioned so that a portion of each conduit is oriented between consecutive upper surfaces and a portion is positioned between consecutive lower surfaces, in a C-shaped traveling loop bounded above by the upper restraints and below by the lower restraints. The singular C-shaped loop of each cable passes across the gap between the aligned vertical surfaces upon relative motion of the two members. The cables are substantially stationary at opposite portions with respect to each of the respective members, either terminating at a junction or affixed to the surfaces. By forming the surfaces of suitable metallic or other material, structural integrity and electrical separation is provided by the surfaces. A number of spring type devices are affixed to the upper surfaces so as to support the upper portion of each looped cable. The cables between any two consecutive aligned surfaces are preferably of substantially the same diameter and the spring device has an upper section and a crimped lower segment. The lower segment is spaced from one of the vertical surfaces a distance less than the diameter of the cables, while the upper section is spaced therefrom a distance greater than the common diameter. The upper portion of the cables can therefore be supported above the crimped section between the spring device and a mating vertical surface. On relative motion of the members, a portion of the cables slide from between one pair of vertical surfaces to between the mating pair of surfaces, the C-shaped loop sliding with respect to both members. Where one member is fixed and the other moves, the moving vertical surfaces acquire additional cable as motion is in the direction which appears to pull cable from the stationary surfaces. Conversely, the stationary surfaces acquire cable as the moving surfaces rotate or move rectilinearly in the opposite direction. The spring device flexes laterally during these motions so that the cables pass freely into and out of contact with the springs. The apparatus can advantageously include a number of cables between the same vertical surfaces, each cable being positioned within the vertical area enclosed by an adjacent outer cable. With cables of similar diameter, the lateral spacing between adjacent vertical surfaces is then slightly larger than this diameter, and the vertical gap between sets of surfaces is smaller than the diameter to alleviate the potential for cable binding upon motion. |
abstract | The machinery system and its application herewith, intended to create an opening in the roof of a Reactor Primary Containment of Generation II and III Nuclear Power Electric Generating Station. This opening is necessary to replace an aging nuclear reactor with a new, safer and more efficient reactor. Generation II and III Nuclear Power Electric Generating Stations include General Electric (GE) Boiling Water Reactor BWR/2, 3, 4, 5 and BWR/6 located in Mark II, and Mark III wet containments and Pressurized Water Reactors manufactured by Westinghouse, Combustion Engineering and Babcock and Wilcox located in dry containments. Until this time, existing reactor replacement was not possible due to Reactor Primary Containment structural enclosure configuration. The Dual Head Vertical Milling Machine System will remove a Reactor Primary Containment Dome Segment thus providing an opening, allowing reactor replacement and the electric generating station to remain operational for an other 40 years and beyond. Original containment integrity will be reinstated by closure of the opening. |
|
description | The present invention relates to a method of imaging a phase change of a radiation and a radiation imaging apparatus thereuse. As opposed to a related-art method of imaging an absorption contrast of a radiation, in particular, an X-ray, a method of imaging a phase change has been attracting attention for its excellent property of rendering a light element, that is, a soft tissue of a living body and the like. In particular, among others, a method in which an X-ray Talbot-Lau interferometer is used has been researched and developed for practical use because a general X-ray source used in medical settings and laboratories may be used therein. In a related-art Talbot-Lau interferometer, an absorption grating is arranged at an image forming position of a self-image that is an interference pattern based on the Talbot effect, which is formed by a phase grating and has a period of several micrometers, and Moire fringes, which are generated as a result and form a pattern having a period that is larger than that of the self-image, are imaged by a two-dimensional radiation detector. The reason why the Moire fringes are imaged in the method using the X-ray Talbot-Lau interferometer as described above is that there is no detector capable of directly resolving the self-image. Moreover, the absorption grating is produced by a semiconductor process with silicon and the like, and gold plating and the like, and hence is very expensive. To address this problem, in Japanese Patent No. 4702455, there is disclosed a unit configured to form Moire fringes using, instead of an absorption grating that acts on an X-ray, an optical diffraction grating that acts on visible light, which is obtained by converting the X-ray once by a scintillator. This unit has an advantage of eliminating the need for an absorption grating, which uses gold and is expensive, and a feature that the two-dimensional radiation detector may be formed of the following three components: a scintillator, an optical diffraction grating, and an image pickup element. However, also in the above-mentioned two-dimensional radiation detector, the two-dimensional radiation detector is formed of the three components: the scintillator, the optical diffraction grating, and the image pickup element, and hence there has been a need to separately produce the optical diffraction grating in addition to the scintillator and the image pickup element. According to one embodiment of the present invention, there is provided a radiation phase change detection method for detecting a phase change of a radiation, the radiation phase change detection method including changing an interference state of the radiation using a phase grating configured to cause interference in the radiation radiated by a radiation source, a scintillator configured to convert the radiation into light, and an image pickup element including two-dimensionally arranged pixels (hereinafter also referred to as “two-dimensional optical image pickup element”), in which as the image pickup element, an element is used which is incapable of sampling a period of a self-image of the radiation generated through the phase grating, and is capable of sampling interference fringes generated between the period of the self-image and a period of a pixel pitch of the two-dimensional optical image pickup element. According to one embodiment of the present invention, there is provided a radiation phase change detection method for detecting a phase change of a radiation, the radiation phase change detection method including: arranging a two-dimensional optical image pickup element, which includes a scintillator, so that, when a period of a self-image generated through a phase grating is defined as D1, and a pixel pitch of the two-dimensional optical image pickup element is defined as D2=k*D1, k falls in a range of ½<k≤3/2, and so that interference fringes formed by D1 and D2 depending on a relationship in arrangement of the two-dimensional optical image pickup element with respect to the self-image have a period of 2 times D2 or more and 100 times D2 or less; acquiring a phase change of the interference fringes before and after insertion of an object; and outputting an image on a phase change of the radiation caused by at least the insertion of the object. Further, according to one embodiment of the present invention, there is provided a radiation imaging apparatus, which is a radiation imaging apparatus utilizing the above-mentioned radiation phase change detection method. According to the radiation phase change detection method of the present invention, interference fringes can be generated and a phase change of the interference fringes can be detected with the structure including only the scintillator and the image pickup element without the need for an absorption grating or an optical diffraction grating, which has been required in the related-art method. Further features of the present invention will become apparent from the following description of exemplary embodiments with reference to the attached drawings. Now, modes for embodying the present invention are described with reference to the drawings. In FIG. 1, an example of the structure of a radiation imaging apparatus based on a radiation phase change detection method according to the present invention is illustrated taking, as an example, a case of using an X-ray as a radiation. The radiation imaging apparatus according to the present invention includes, as main components, an X-ray source 11, a source grating 12, a phase grating 13, a scintillator 15, and a two-dimensional optical image pickup element 16. As the X-ray source 11, tubes of different target materials may be selected in consideration of characteristics of an object 14 and the like. Examples of the target material include Cu, Ag, Mo, and W. When a general X-ray source 11 having a large focal point size is used, the source grating 12 may be inserted to adjust a size of a radiation source. As the source grating 12, such source grating 12 as described in Japanese Patent Application Laid-Open No. 2010-249533 may be used. Moreover, the phase grating 13, which may be produced through a semiconductor process and is made of Si or the like, plays a role of causing interference between spherical waves diffracted by respective slits of the phase grating 13 to form a self-image at a point corresponding to the Talbot length downstream of a grating surface. In the radiation phase change detection method according to the present invention, how the self-image is changed with a change in interference state of the radiation before and after insertion of the object 14 is captured to detect a phase change of the self-image. The object 14 may be inserted between the phase grating 13 and the X-ray source 11 or the source grating 12, or between the phase grating 13 and the scintillator 15. In each of FIG. 2A and FIG. 2B, there is illustrated an example of the structure of a radiation imaging apparatus based on a radiation phase change detection method according to the related art. Also in the radiation phase change detection method according to the related art, as in the radiation phase change detection method according to the present invention, a change of the self-image caused by insertion of the object 14 is captured to detect the phase change of the self-image. The radiation imaging apparatus in each of FIG. 2A and FIG. 2B includes an X-ray source 11, a source grating 12, a phase grating 13, a scintillator 15, a two-dimensional optical image pickup element 16, and an absorption grating 21 (FIG. 2A) or an optical diffraction grating 22 (FIG. 2B). As with the phase grating 13, the absorption grating 21 is a grating for the X-ray having a very high aspect ratio, which is formed by producing a mold made of Si or the like through the semiconductor process, and further being implanted through, for example, plating using an element having a high atomic number, for example, gold. In particular, having a periodicity of 10 μm or less and requiring a thickness of 100 μm or more, the absorption grating 21 is a very expensive part in consideration of a process cost and a cost of materials. In the radiation phase change detection method according to the related art, the self-image having a period of 10 μm or less cannot be directly resolved. Therefore, as illustrated in FIG. 2A, there is adopted a method in which interference fringes of the X-ray, which have a period that is larger than those formed by the self-image and the absorption grating 21, are converted into visible light by the scintillator 15, and are imaged by the two-dimensional optical image pickup element 16. Moreover, in consideration of responsiveness for sufficiently absorbing the X-ray and radiating a large amount of scintillation light, the scintillator 15 needs to have a thickness of at least 150 μm or more depending on energy of the X-ray. Moreover, in FIG. 2B, there is illustrated a radiation phase change detection method in which, instead of the absorption grating 21 for the X-ray, the self-image of the X-ray is converted into the visible light by the scintillator 15, and then the interference fringes are generated with the optical diffraction grating 22. In the radiation phase change detection method illustrated in FIG. 2B, a scintillator 15 having a very high definition is required. Therefore, the scintillator 15 needs to have a very small thickness, and it is assumed that the radiation phase change detection method illustrated in FIG. 2B has significantly reduced detection sensitivity as compared to the radiation phase change detection method illustrated in FIG. 2A. Moreover, it is also required to produce the optical diffraction grating 22, although there is no need to use gold. In contrast, in the radiation phase change detection method according to the present invention, the absorption grating 21 or the optical diffraction grating 22 is not required. The reasons are as follows. The first reason is that a scintillator 15 capable of resolving 100 lp/mm (lp: line pair) at least with a thickness of 150 μm is used. The expression “100 lp/mm” means that it is spacial frequency with 100 light and dark lines per mm, and the scintillator 15 has a very high resolution which enables discrimination of light having a periodicity of 10 μm or less. The second reason is that a light receiving portion of the two-dimensional optical image pickup element 16 has a predetermined pixel pitch, and is capable of causing interference with the self-image. Therefore, the interference fringes may be generated, and desired imaging may be performed without inserting the optical diffraction grating 22 as in the radiation phase change detection method illustrated in FIG. 2B. In other words, the two-dimensional optical image pickup element 16 is incapable of sampling the period of the self-image of the radiation generated through the phase grating 13, but is capable of sampling the interference fringes generated through the phase grating 13. In the present invention, it is preferred that the above-mentioned interference be Moire interference, and that the above-mentioned interference fringes be Moire fringes. For the two reasons described above, in the radiation phase change detection method according to the present invention, when it is defined that D2=k*D1, where D1 represents a period of the self-image formed by the phase grating 13, and D2 represents the pixel pitch of the two-dimensional optical image pickup element 16, and when k falls in a range of ½<k≤3/2, interference can be caused between the self-image and the light receiving portion of the two-dimensional optical image pickup element 16. Moreover, it is preferred that a relationship in arrangement of the two-dimensional optical image pickup element 16 with respect to the self-image be adjusted so that the interference fringes have a period of 2 times the pixel pitch D2 or more and 100 times the pixel pitch D2 or less. An example of the adjustment of the relationship in arrangement of the two-dimensional optical image pickup element 16 with respect to the self-image may include adjustment of a rotation angle of the two-dimensional optical image pickup element 16 with respect to the self-image. When the period of the interference fringes is larger than 100 times the pixel pitch D2, the resolution of the obtained image becomes much lower with respect to the pixel pitch D2, and the two-dimensional optical image pickup element 16 has an unnecessarily high resolution with respect to the period of the interference fringes. In this case, the phase change of the radiation caused by the object 14 may be detected with the radiation phase change detection method according to the present invention, but the number of light beams that enter each pixel of the two-dimensional optical image pickup element 16 is significantly reduced, and may cause statistical noise. Therefore, it is preferred to adjust the rotation angle of the two-dimensional optical image pickup element 16 so that the interference fringes have the period of 2 times the pixel pitch D2 or more and 100 times the pixel pitch D2 or less. It is more preferred that the interference fringes have a period of 2 times the pixel pitch D2 or more and 20 times the pixel pitch D2 or less, and it is further preferred that the interference fringes have a period in the vicinity of 4 times the pixel pitch D2. As shown in Table 1, a relationship between the period of the interference fringes when the self-image has a period of 8 μm and the pixel pitch D2 was calculated with Expression 1. When the pixel pitch D2 has a value close to the period D1 of the self-image, the period of the interference fringes is significantly increased. In this case, as calculated with Expression 2, a rotation angle θ of the two-dimensional optical image pickup element 16 may be adjusted to adjust the interference fringes to have the period of 2 times the pixel pitch D2 or more and 100 times the pixel pitch D2 or less (Table 1). For example, when the self-image has a period of 8 μm, and a pixel pitch is 8.01 μm, the rotation angle θ of the two-dimensional optical image pickup element 16 may be rotated by 12° to adjust the period of the interference fringes from 800 times the pixel pitch D2 to 4.78 times the pixel pitch D2. The two-dimensional optical image pickup element 16 may be rotated clockwise or counterclockwise, and the rotation angle may be adjusted arbitrarily so that the interference fringes have the period of 2 times the pixel pitch D2 or more and 100 times the pixel pitch D2 or less. Further, the two-dimensional optical image pickup element 16 may be adjusted by being translated.Period of interference fringes=D1×D2/|D1−D2| (1)Period of interference fringes=D={(1/D1)+(1/D2)−((2/D1*D2)*cos θ)}^(−½) (2) TABLE 1Period of InterferencePeriod D1 ofPixel PitchFringes/Pixel Pitch D2Self-imageD2WithoutRotated by[μm][μm]Adjustment12°8.0012.002.00—8.0010.004.002.928.009.008.003.938.008.01800.004.788.007.008.004.318.006.004.003.248.004.012.01— In the relationship: D2=k*D1 in the radiation phase change detection method according to the present invention, when k is ½ or less (k≤½), the period of the self-image may be directly measured instead of the period of the interference fringes based on the sampling theorem. The radiation phase change detection method according to the present invention provides an advantage that an element having a larger pixel pitch D2 may be used by forming the interference fringes. For example, in an optical image pickup element (e.g., CCD or CMOS image sensor) having a relatively large area, when the pixel pitch D2 is small, a pixel number is increased, and there are problems in that reading takes time, and in that yield of the sensor itself is affected. Therefore, in the radiation phase change detection method according to the present invention, the pixel pitch D2 is set as large as possible in a desired range of image resolutions to have an effect that the above-mentioned problems can be solved. For example, in the radiation phase change detection method according to the present invention, when the period of the self-image is 10 μm, a two-dimensional optical image pickup element 16 including pixels having a pixel pitch D2 of 5 μm or more may be used. Meanwhile, when the pixel pitch D2 does not satisfy k≤3/2, a ratio of the pixel pitch D2 to the period of the interference fringes becomes smaller than 2 times, and the interference fringes themselves cannot be sampled. Therefore, in the radiation phase change detection method according to the present invention, when the period of the self-image is 10 μm, a two-dimensional optical image pickup element 16 including pixels having a pixel pitch D2 of 15 μm or less may be used. In the above description, there has been described the example in which the X-ray is used as the radiation, but any radiation may be used as long as the radiation may excite and cause the scintillator 15 to emit light. Examples of such radiation include a γ-ray and a neutron ray. Moreover, a two-dimensional optical image pickup element 16 having any aperture ratio in one pixel, that is, any ratio between an area of a light receiving portion 32 and an area of a pixel may be used, but the aperture ratio is reflected on a contrast of fringes obtained when the interference fringes are imaged. In FIG. 3, the two-dimensional optical image pickup element 16 having a pixel pitch 33 is illustrated. When an aperture ratio (fill factor: FF) expressed by P/S, where S represents an area of a pixel 31 of the two-dimensional optical image pickup element 16, and P represents an area of the light receiving portion 32 in the pixel 31, falls in a range of from 0.01 to 1.0, a satisfactory contrast of the interference fringes may be obtained. For example, when the self-image had a period of 8 μm, and the pixel pitch 33 of the two-dimensional optical image pickup element 16 was 7.2 μm, the aperture ratio was changed to obtain images. When the aperture ratio was 0.56, an image of the interference fringes shown in FIG. 4A was obtained, and when the aperture ratio was changed, a cross-sectional profile of the image was changed as shown in FIG. 4B. As shown in FIG. 4B, an amplitude (contrast) of the interference fringes is changed with the change in aperture ratio, but the periodicity is not impaired. Therefore, a wide range of aperture ratios may be selected. In the radiation phase change detection method according to the present invention, a phase change in a result obtained by imaging the interference fringes generated between the self-image and the two-dimensional optical image pickup element 16, which is caused by presence or absence of the object 14, is analyzed. An analysis method may be any method, and it is preferred that a method of analyzing the phase change be a Fourier transform method. The Fourier transform method may be used to analyze the phase change at a resolution corresponding to a period of the obtained interference fringes. Moreover, the phase change may be analyzed by fringe scanning interferometry, with which a high resolution may be obtained although a plurality of imaging operations are required. In the radiation phase change detection method according to the present invention, the phase change of the interference fringes generated between the self-image and the two-dimensional optical image pickup element 16 by the insertion of the object 14 may be acquired as an image through the above-mentioned analysis. Therefore, in the radiation phase change detection method according to the present invention, the phase change of the radiation caused by the insertion of the object 14 may be detected by comparing an image of the interference fringes obtained when the object 14 is not inserted and an image of the interference fringes obtained when the object 14 is inserted with each other to output a distortion in the interference fringes due to a change in period D1 of the self-image, which is caused by the phase change of the radiation, as an image. [Eutectic Phase-Separated Scintillator] In FIG. 5, an example of the scintillator 15 used in the radiation phase change detection method according to the present invention is illustrated. It is preferred that the scintillator 15 used in the radiation phase change detection method according to the present invention be an uniaxial eutectic phase-separated scintillator 50, which includes a plurality of first phases 51, which serves as a scintillator, and a second phase 52, has uniaxiality in a direction connecting two surfaces (hereinafter sometimes referred to as “first surface” and “second surface”) that are not on the same plane, and has a light guiding property in that direction. The plurality of first phases 51 are phases having columnar fiber structures surrounded by the second phase 52 and each having a unidirectional property along an optical waveguide direction 53, and exist at a certain average structural period 55 in the second phase 52. It is preferred that the first surface and the second surface be substantially perpendicular (within ±20°) to the direction (optical waveguide direction 53) along which the first phases 51 extend. Moreover, the first phases 51 need to be phases having a high refractive index, and the second phase 52 needs to be a phase having a low refractive index. Further, the first phases 51 serve as a scintillator, and hence the first phases 51 are capable of being excited by absorbing the radiation, and of guiding emitted light as with an optical fiber. As materials of the first phases 51 and the second phase 52 of such eutectic phase-separated scintillator 50, a material system of such combination as shown in Table 2 may be used. TABLE 2Material of FirstMaterial of SecondPhases 51Phase 52GdAlO3Al2O3CsINaNO3ZnOZnF2 A shape of the first phase 51 is not limited to the cylindrical shape as illustrated in FIG. 5, but may be a columnar body having an elliptical cross section, a columnar body having a side surface that is formed of an arbitrary curved surface, or a columnar body formed of a plurality of surfaces and having a polygonal cross section. It is preferred that a diameter 54 of a column of the first phase 51 be 50 nm or more and 30 μm or less. In other words, when the first phase 51 has a circular cross section, it is preferred that a diameter of the circular cross section be 50 nm or more and 30 μm or less. Moreover, when the first phase 51 has an elliptical cross section, it is preferred that a major axis and a minor axis of the elliptical cross section be 50 nm or more and 30 μm or less. Further, when the first phase 51 has a polygonal cross section, it is preferred that, of values that are twice distances between a center of gravity and vertices, the smallest value and the largest value be 50 nm or more and 30 μm or less. Moreover, it is preferred that the average structural period 55 of the first phases 51 in the eutectic phase-separated scintillator 50, which is expressed by an average distance between centers of gravity of cross sections of the first phases 51, be 500 nm or more and 50 μm or less. It should be noted, however, that the average structural period 55 needs to be smaller than half the period D1 of the self-image. The first phases 51 may have an interruption, a branch, coupling, and the like, but it is preferred that the first phases 51 be continuous between the first surface and the second surface. A thickness 56 of the eutectic phase-separated scintillator 50 may be selected as appropriate depending on energy of the absorbed radiation so that the first phases 51 serve as a scintillator. It is preferred that the thickness 56 of the eutectic phase-separated scintillator 50 for use in the radiation phase change detection method according to the present invention be 10 μm or more and 40 mm or less, and it is more preferred that the thickness 56 be 50 μm or more and 5 mm or less. The eutectic phase-separated scintillator 50 having such structure has performance capable of sufficiently resolving the period D1 of the self-image, which is 8 μm, when the average structural period 55 is 1.68 μm, the first phase 51 is GdAlO3 having columns of an average diameter 54 of 1.24 μm, and the thickness 56 of the eutectic phase-separated scintillator 50 is 150 μm, for example. In FIG. 6, there is shown a transmission image obtained by allowing a visible radiation simulating a self-image having a period of 8 μm to enter from the rear surface to be transmitted by the above-mentioned eutectic phase-separated scintillator 50. As can be seen from FIG. 6, even the self-image of 10 μm or less is transmitted by each of the first phases 51. Therefore, information on the self-image is not significantly modulated or a diffraction phenomenon is not caused, and hence the self-image is not degraded, due to the structure of the eutectic phase-separated scintillator 50. Therefore, when the period D1 of the self-image and D2, which is the pixel pitch 33 of the two-dimensional optical image pickup element 16, satisfy the relationship in the radiation phase change detection method according to the present invention, it is possible to provide a radiation phase change detection method, which eliminates the need for the absorption grating 21 or the optical diffraction grating 22 for generating the interference fringes. In the radiation phase change detection method according to the present invention, it is preferred to adopt an arrangement in which the first phases 51 of the eutectic phase-separated scintillator 50 are brought into direct contact with an image pickup surface of the two-dimensional optical image pickup element 16. In this case, the first phases 51 may be bonded to the image pickup surface of the two-dimensional optical image pickup element 16. The two-dimensional optical image pickup element 16 may be a special element, or a general element in which a microlens is provided for each pixel. Moreover, even in the case of direct bonding, there is no need to bring a light receiving surface and the scintillator 15 into direct contact with each other, and various protective layers may be interposed. Further, in order to take measures to reduce spot noise generated by the radiation that has been transmitted through the first phases 51 of the eutectic phase-separated scintillator 50 entering the two-dimensional optical image pickup element 16, it is preferred to insert, between the eutectic phase-separated scintillator 50 and the two-dimensional optical image pickup element 16, a fiber optic plate (hereinafter also referred to as “FOP”) having a periodicity that is half the period of the self-image or less and no effect on a periodic component of interest. [Radiation Phase Change Detection Apparatus] The radiation imaging apparatus according to the present invention includes a radiation source, a phase grating 13, a scintillator 15, and a two-dimensional optical image pickup element 16. In the phase change detection method according to the present invention, the need for the absorption grating 21 or the optical diffraction grating 22 is eliminated, and hence there is no need to provide such component in the radiation imaging apparatus according to the present invention. The radiation imaging apparatus according to the present invention may also include a driving system configured to adjust a position of each part and insert the object 14, an arithmetic section configured to extract and analyze a phase change, and an image acquisition unit configured to visualize the phase change. When the X-ray is used as the radiation, the radiation imaging apparatus may be formed using the X-ray source 11, the phase grating 13, the scintillator 15, and the two-dimensional optical image pickup element 16, which have been described above as components of the radiation phase change detection method according to the present invention. In this case, an apparatus length of the radiation imaging apparatus may be 2 m or less. Using a eutectic phase-separated scintillator 50 in which the material of the first phases 51 is GdAlO3 and the material of the second phase 52 is Al2O3, an examination was performed on whether or not the information on the self-image was significantly modulated by the eutectic phase-separated scintillator 50. The above-mentioned eutectic phase-separated scintillator 50 is a eutectic phase-separated scintillator 50 that is obtained by adding Tb to Gd sites and emits green light, for which the two-dimensional optical image pickup element 16 has high receiving sensitivity. Moreover, the above-mentioned eutectic phase-separated scintillator 50 has a thickness 56 of 150 μm and an average structural period 55 of 1.68 μm, and includes the first phases 51 having columns of an average diameter 54 of 1.24 μm. As an X-ray optical system, the X-ray source 11 containing tungsten as a target and having radiation quality of 28 keV with a 1-mm Al filter was used with an insertion of the source grating 12 having a pitch of 12.75 μm. A period D1 of the self-image formed by a two-dimensional phase grating 13 made of silicon was 8.24 μm. Under the above-mentioned conditions, the eutectic phase-separated scintillator 50 was arranged at a position at which the self-image was formed, and the self-image was directly taken for about 40 seconds while being enlarged by a lens. At this time, a distance from the source grating 12 to the eutectic phase-separated scintillator 50 is 1,185 mm, and a resolution per pixel in a lens optical system is 0.65 μm. A self-image of an exit surface of the eutectic phase-separated scintillator 50 taken under the above-mentioned conditions is shown in FIG. 7. In FIG. 7, the self-image having a period of 8.24 μm was clearly rendered. As a result, it was able to be confirmed that a periodic pattern of the self-image was not modulated by the eutectic phase-separated scintillator 50. Further, an image was taken by changing the thickness 56 of the eutectic phase-separated scintillator 50 to 300 μm and bonding the eutectic phase-separated scintillator 50 to a CCD having D2, which is the pixel pitch 33, of 4.4 μm (k=0.534 based on the expression: D2=k*D1 in the radiation phase change detection method according to the present invention). Then, the period D1 of 8.24 μm of the self-image was detected as interference fringes (Moire fringes) having a period of about 9.4 μm. Moreover, a change in Moire fringes before and after insertion of a nylon ball having a diameter of 4 mm, which was used as the object 14, was able to be extracted and analyzed as modulation of the frequency component of interest with the Fourier transform method. As described above, it was confirmed that the radiation phase change detection method according to the present invention is effective as a method of detecting the phase change in the period of the interference fringes caused by the interference between the self-image, which is formed by the radiation and the phase grating 13, and the two-dimensional optical image pickup element 16 without degrading the self-image by the scintillator. An image of the interference fringes was taken as in Example 1 except that the thickness 56 of the eutectic phase-separated scintillator 50 in Example 1 was changed to 250 Tim, that an FOP having a thickness of about 2.5 mm and a periodicity of about 3 μm was inserted between the eutectic phase-separated scintillator 50 and the two-dimensional optical image pickup element 16, and that imaging time was changed to 8 seconds. As a result of performing analysis with the Fourier transform method, as shown in FIG. 8A, the period D1 of the self-image was detected as interference fringes (Moire fringes) having a period of about 9.4 μm though with diffraction spots having a long period caused by the FOP. In addition, a phase change before and after insertion of a nylon ball having a diameter of 4 mm was extracted and analyzed with the Fourier transform method. As a result, an image of FIG. 8B was obtained, and it was able to be confirmed that, in the radiation phase change detection method according to the present invention, the insertion of the FOP having a periodicity of half the period D1 of the self-image or less did not have a problematic effect on the periodic component of the period D1 of the self-image in the radiation phase change detection method according to the present invention, and had the effect of reducing spot noise. With the radiation phase change detection method according to the present invention, the need for the absorption grating or the optical diffraction grating is eliminated, and hence the radiation phase change detection apparatus can be formed in a compact size. The radiation phase change detection method according to the present invention is applicable to the X-ray, the γ-ray, and the neutron ray. In particular, in an X-ray region, the radiation phase change detection method can be used for phase change imaging apparatuses for medical and industrial purposes. While the present invention has been described with reference to exemplary embodiments, it is to be understood that the invention is not limited to the disclosed exemplary embodiments. The scope of the following claims is to be accorded the broadest interpretation so as to encompass all such modifications and equivalent structures and functions. This application claims the benefit of Japanese Patent Application No. 2016-091055, filed Apr. 28, 2016, which is hereby incorporated by reference herein in its entirety. |
|
summary | ||
044420668 | summary | BACKGROUND OF THE INVENTION 1. Field of the Invention The invention concerns a supporting floor for the core of a nuclear reactor. More particularly, the supporting floor comprises a plurality of graphite blocks arranged in vertical columns. The nuclear reactor, preferably a gas cooled nuclear reactor, comprises a reactor core in the form of a pile of spherical fuel elements surrounded by a cylindrical side reflector and a thermal side shield arranged at a distance from the side reflector. The vertical columns of the supporting floor are provided with bores in the longitudinal direction and rest on the base of the nuclear reactor itself. 2. Background of the Prior Art West German Offenlegungsschrift No. 1 956 226 discloses a gas cooled nuclear reactor having a core of graphite blocks arranged in vertical columns. The core is supported on the floor of the reactor pressure vessel by means of several layers of a refractory material, for example, graphite. Each column of the core is supported by its own supporting column. The supporting columns consist of several graphite blocks stacked on top of each other. Access for the cooling gas to the core is provided by means of channels formed into the graphite blocks of the support layers and connected with gas conduits in the core and with a gas space outside the core. The entire structure of supporting columns is surrounded by packing blocks arranged between the structure and the lining of the pressure vessel. West German Published Application No. 1 261 606 shows a nuclear reactor with a radiation reflector inserted between the thermal shield and the core of the reactor. The reflector consists of a lateral reflector enclosing a cylindrical cavity and a lower reflector serving as the supporting floor. The lower reflector is formed by two layers of ashlar shaped graphite blocks differing in their longitudinal direction from one layer to the other. In each of the layers the blocks arranged around a central opening are maintained together and locked against each other by means of wedges. The lateral reflector, which consists of graphite blocks stacked in layers, one upon the other, is supported by the thermal side shield. A plurality of elastic structural parts acting in the radial direction provides this support. The prior art further includes a supporting floor for a pebble bed reactor consisting of a gas permeable supporting layer of spheres of high temperature resistant material and a support structure designed for the weight of the supporting pebble bed layer and the fuel elements. The fuel elements are piled directly onto the supporting pebble layer. A layer of high temperature resistant tiles is arranged between the support structure and the supporting pebble layer. In another pebble bed reactor, the THTR-300 MWe, the support floor for the pile of fuel pellets comprises a plurality of hexagonal graphite blocks arranged in freely movable columns and containing axial bores for the cooling gas. Each column formed by the graphite blocks is supported individually by a circular column. The circular column is built into a floor of graphite plates. The expansion gaps obtained by a reduction of the nominal dimensions of the graphite blocks permit the unimpaired thermal expansion of the elements of the support floor without exceeding the overall dimensions. Under certain non-stationary operating conditions, or in the case of failures, respectively, these gaps may add up and relatively large individual gaps may be formed. However, in view of the dimensions of the THTR-300 MWe, the restoring forces required to close the gaps are of negligible importance. A supporting structure for a pebble bed reactor of greater capacity is known from West German Offenlegungsschrift No. 27 18 493. This support structure consists of several layers of prismatic graphite blocks arranged over each other and built up as a closed unit without expansion gaps. The blocks of one layer are keyed into the blocks of the adjacent layer. The preferred hexagonal graphite blocks display widths across their flats in the upper layers that are different from those in the bottom layer, which is formed of a plurality of supporting units. Each supporting unit is resting in its central area of a circular column and is composed of a number of carrier segments. The support structure represents a stable and rigid plate. Because of the keying of the prismatic blocks, no restoring force is required. However, in the case of this known structure, certain specific measures must be taken to counter the deformations occurring in the bottom layers of the nuclear reactor or in the bottom of the pressure vessel, respectively. SUMMARY OF THE INVENTION The present invention is an improvement of the abovedescribed state of the art and is directed specifically to providing a support floor for a nuclear reactor having spherical fuel elements. The structure of the present invention avoids the reactive forces and the excessive thermal stresses often found in the known reactors and generated as the result of the existance of radial temperature profiles. Adjustments without strain to the deformations of the bottom of the pressure vessel are also possible in the structure of this type. These and other objects are attained according to the invention by placing the vertical columns as independent individual columns, directly adjacent to each other, without expansion gaps and by maintaining the columns together in the horizontal direction by means for retaining the columns in their original relative horizontal position. This means is arranged in the annular space between the lower part of the side reflector and the thermal side shield. The means is preferably a plurality of restoring elements such as spring supports or support struts. The supporting floor arrangement of the invention performs all of the functions required of it. For example, it transmits the vertical forces of the pile of fuel elements downward, it passes the horizontal forces by way of restoring elements to the thermal side shield, it also performs the task of shielding and of conducting the hot gas from the reactor core to the hot gas collector space. Because the supporting floor has been resolved into a plurality of independent, individual columns, the differential thermal expansions of the columns are not restricted. Stationary and non-stationary temperature fields, therefore, do not lead to the generation of reactive forces or substantial thermal stresses. The support floor is not sensitive relative to manufacturing and assembly tolerances and adapts without strain to the deformations of the bottom of the pressure vessel. Vibrations of the individual columns of the support floor are practically excluded, because the support floor of a pebble bed reactor customarily has an inclined surface (for the removal of the spherical fuel elements at least one pebble outlet tube must be provided with a conical tube inlet located within the support floor) and the columns are maintained together by means of elements that retain, restore or reset the columns under stress. The vertical columns of the support floor according to the invention preferably have a hexagonal cross section. The columns are placed adjacent to each other so that they are generally in contact with each other. (Of course some gaps due to manufacturing conditions are possible and tolerable). Advantageously, an additional circular column may be provided for each hexagonal column. The circular column rests on the bottom layers of the nuclear reactor while the hexagonal column rests on the circular column. The circular columns conveniently have diameters smaller than the width of the bottom surface of the hexagonal columns. The free space created in this manner around the circular columns forms a hot gas collector space of the nuclear reactor. The circular columns acting as linked supports for the hexagonal columns, may be built into the bottom layers of the nuclear reactor. In high capacity nuclear reactors having large reactor cores, the retaining means is comprised of a plurality of restoring elements. These are advantageously designed in the form of spring supports. By such means, the vertical columns are maintained together in the radial direction so that no gaps or only acceptably small gaps, exist between the vertical columns. Preferably, the spring supports are arranged so that the gaps appearing between the vertical columns after extended operating periods or following a non-stationary operation are limited to a predetermined maximum size. In the process, the horizontal forces of the pile of fuel elements must be safely transferred to the thermal side shield. At the same time, the differential thermal expansions of the support floor and the thermal side shield must also be possible. In nuclear reactors with lesser capacities, support struts may be used as the restoring elements; they must be arranged with clearance between the support floor and the thermal side shield. The differential radial thermal expansions of the support floor and the thermal shield are taken into account by that the maximally possible differential thermal expansion of the above-mentioned two structural parts are set at the support struts as clearances. When the support floor according to the invention is used in nuclear reactors having their reactivities affected by means of absorber balls with substantially smaller diameters than those of the spherical fuel elements, spring supports are utilized (independently of the size of the core structure) as the restoring elements. The restoring force is designed so that the dimensions of the gaps being formed between the vertical columns remains smaller than the diameter of the absorber balls, even toward the end of the operation of the nuclear reactor. Gaps are created not only by the temperature load during an extended operating period, but occur also under certain operating conditions. Thus, gaps are dependent on the operation and temperature conditions. The restoring forces of the spring supports have the function of suppressing the gaps between the vertical columns or at least to limit their size, in order to prevent the penetration of the absorber balls into any gaps that form and thus causing the straining of the assembly. It may be appropriate to provide a number of gaps in the side reflector. This will prevent the side reflector from becoming unable to transmit forces to the support floor due to a pressure ring support effect, under certain conditions. According to an advantageous further embodiment of the invention, the bores for the cooling gas in the vertical columns are provided in relation to number, diameter and distance so that in the individual vertical columns either none or only very slight non-stationary thermal stresses are generated. Numerous individual bores may be arranged to pass through each vertical column in the longitudinal direction such that the temperatures prevailing in the support floor arrangement adapt uniformly to temperature transients, whenever such occurs. |
claims | 1. A glove box comprising:an enclosure having a working chamber defined therewithin;said enclosure having a viewing portion;said enclosure having at least one gauntlet extending into the working chamber;piping operable to permit a selected atmosphere to be established within said chamber;an access by which to introduce a work-piece into said chamber; andsaid enclosure having an angular adjustment by which to tilt at least one of said gauntlet and said viewing portion. 2. The glove box of claim 1 wherein said access include an environmental lock antechamber. 3. The glove box of claim 2 wherein said environmental lock antechamber has piping to permit said antechamber to be flushed and charged with gases independently of said chamber. 4. The glove box of claim 1 wherein said glove box has at least a first parameter read out display, said read-out display being located outside said chamber. 5. The glove box of claim 1 wherein a seat is mounted within said chamber in which to accommodate the work-piece. 6. The glove box of claim 5 in which said enclosure is angularly adjustable relative to said seat. 7. The glove box of claim 1 having at least a first sealed utility penetration through a wall of said enclosure. 8. The glove box of claim 1 wherein said glove box has a heat exchanger mounted therewithin by which to adjust temperature within said chamber. 9. The glove box of claim 1 wherein said glove box has a tool interface therewithin, and said tool interface includes a coolant line connection. 10. The glove box of claim 1, said at least one gauntlet including a first gauntlet and a second gauntlet, wherein said glove box includes:an independently flushable access antechamber mounted to said enclosure;a seat in which to mount a work-piece;a welding electrode holder in which to mount a consumable electrode;a welding electrode power connection that passes through a sealed penetration of said enclosure, in use said welding electrode holder being mounted thereto;a heat exchanger mounted within said enclosure, and operable to extract heat therefrom;controls of at least one of (a) power to said welding electrode holder; and (b) cooling of said heat exchanger. 11. The glove box of claim 1 comprising:piping by which to introduce gases into said chamber; andan environmental control system operable to govern temperature within said chamber. 12. The glove box of claim 1 comprising:a tool for use by an operator to engage the work piece; anda controller of said tool mounted within said chamber, said controller being adjustable by an operator wearing said first gauntlet. 13. The glove box of claim 11 wherein said access include an environmental lock antechamber. 14. The glove box of claim 11 wherein a seat is mounted within said chamber in which to accommodate the work-piece; and said enclosure is angularly adjustable relative to said seat. 15. The glove box of claim 11 wherein said glove box has a tool interface therewithin, and said tool interface includes a coolant line connection. 16. The glove box of claim 11, said at least one gauntlet including a first gauntlet and a second gauntlet, wherein said glove box includes:an independently flushable access antechamber mounted to said enclosure;a seat in which to mount a work-piece;a welding electrode holder in which to mount a consumable electrode;a welding electrode power connection that passes through a sealed penetration of said enclosure, in use said welding electrode holder being mounted thereto;a heat exchanger mounted within said enclosure, and operable to extract heat therefrom;controls of at least one of (a) power to said welding electrode holder; and (b) cooling of said heat exchanger. 17. The glove box of claim 12 wherein said access include an environmental lock antechamber. 18. The glove box of claim 12 wherein a seat is mounted within said chamber in which to accommodate the work-piece; and said enclosure is angularly adjustable relative to said seat. 19. The glove box of claim 12 wherein said glove box has a tool interface therewithin, and said tool interface includes a coolant line connection. 20. The glove box of claim 12, said at least one gauntlet including a first gauntlet and a second gauntlet, wherein said glove box includes:an independently flushable access antechamber mounted to said enclosure;a seat in which to mount a work-piece;a welding electrode holder in which to mount a consumable electrode;a welding electrode power connection that passes through a sealed penetration of said enclosure, in use said welding electrode holder being mounted thereto;a heat exchanger mounted within said enclosure, and operable to extract heat therefrom;controls of at least one of (a) power to said welding electrode holder; and (b) cooling of said heat exchanger. |
|
047055774 | abstract | A plate-shaped high power nuclear fuel element containing low enrichment uranium (5 to 20 percent by weight uranium.sup.235 in the uranium component) as fissionable material, the fuel element essentially comprising a UAl.sub.4 plate provided with an aluminum sheath or a sheath of an Al alloy and impurities inherent in the manufacturing process. A process for producing such a UAl.sub.4 plate comprises (a) intimately mixing a powder of low enrichment uranium or uranium compound U.sub.6 Fe (5 to 20 percent by weight U.sup.235) having a particle size in the range from 0.1.mu. to 90.mu. with aluminum powder having a particle size in the range from 0.1.mu. to 100.mu. in a weight ratio range of uranium to aluminum between 1.1 U:1 Al and 2.2 U:1 Al; (b) prepressing the mixture of step (a) at a pressure in the range from 300 MPa to 500 MPa and at room temperature to form a plate-shaped blank; (c) inserting the blank into an Al picture-frame or a picture-frame of an Al alloy and welding it to the frame in vacuo or inert gas argon; (d) rolling the picture-frame in three roll passes, a reduction in thickness of about 1 mm occurring in each of the first and second passes and a reduction in thickness by about 15% occurring during the third roll pass, at a temperature of 800.degree..+-.25.degree. K.; (e) inserting the plate in the frame after the third roll pass between two Mo sheets, inserting the framed plate together with the Mo sheets in a clamping device, and subsequently heat treating the plate at 800.degree..+-.25.degree. K. for a duration of at least 75 hours in order to form UAl.sub.4 in the plate. |
042882893 | summary | BACKGROUND OF THE INVENTION This invention pertains to fusion machines and more particularly to improvements in tokamaks. Tokamaks are machines originally developed by the Russians starting in 1958. After L. A. Artsimovich came to the United States in 1969 and delivered a series of lectures the first American tokamak was built in Princeton in 1970. The goal of these machines is to attain nuclear fusion. It is known (the Lawson criterion) that such fusion can be attained with a net power release when a plasma which is a fifty percent mixture of deuterium and a fifty percent mixture of tritium with a number density (n) is maintained for a time (.tau.) so that the combined product is (5.times.10.sup.14 atoms/cm.sup.3 .times.200 msec=10.sup.14 atoms-sec/cm.sup.3) and held for this time at a temperature of 100.times.10.sup.6 degrees Kelvin. If these two conditions of the number density-time product and the temperature are achieved the amount of electrical energy produced by the fusion of the hydrogen isotopes deuterium and tritium will equal the electrical energy required to produce the plasma. In addition, it is required that heat energy of the plasma be recovered and transformed into electricity at 33% efficiency while the plasma itself is electrically heated at 100% efficiency. It is believed that if the values of the number density-time product and the temperature were increased by only a factor of two or three there would be a large ratio of output power over the input power. The basic design of a tokamak as given by L. A. Artsimovich comprises a stainless steel chamber filled with deuterium gas at low pressure surrounded by a thick copper wall. Surrounding the toroidal chamber are energized windings to induce a toroidal magnetic field B.sub..phi. of about 20 kG within the chamber. Additionally, another set of energized windings induces a magnetic field B.sub.E directed along an axis which is perpendicular to the plane of the toroidal chamber and which passes through the center of curvature of the torus that defines the chamber. Initially the gas is ionized by radiofrequency signals of about 100 kHz at 20 kW. The magnetic field B.sub.E is linearly increased and adjusted at such a rate so that the external electric field is near 0.2 V/cm to minimize the appearance of runaway electrons. This field drives a plasma current that heats the plasma and rises linearly until the plasma current reaches a value of I.sub.P =100 kA and the discharge duration reaches about 100 msec. The plasma is thereby heated to about 6.times.10.sup.6.degree. K. Various modifications and improvements on such machines were made over the years until in 1975 the Alcator tokamak of MIT achieved the values of: n=4.times.10.sup.14 atoms/cm.sup.3 ; and .tau.=18 msec.; n.tau.=0.7.times.10.sup.13 atoms-sec/cm.sup.3 ; and T=10.times.10.sup.6.degree. K. Lately these values have changed such that n and .tau. have increased by about 50% so that n.tau. has doubled but T has decreased by about 30%. Presently, tokamaks are rather close to the desired parameters of n.tau. and T. Some scaling laws suggest that the desired values of n.tau. will be reached with a B.sub..phi. of about 150 kG because the n.tau. product is proportional to B.sub..phi..sup.4. This is about twice the field achieved until now. Such high fields exert tremendous forces. Indeed the 150 kG field exerts a force of 6.5 tons/in.sup.2. In the Alcator tokamak the highest fields achieved are about 85 kG which exerts a force 1/3 of this value. These high fields are technologically achievable, but very expensive to generate. Another problem with present tokamaks is that the plasma pressure contained is somewhat less than 1% of the magnetic pressure. Thus 2 tons/in.sup.2 of magnetic pressure contain a plasma pressure of 20 lbs/in.sup.2. This is bad not only because of the high capital costs of the magnetic fields, but also because at these low values of .beta. (ratio of plasma pressure to magnetic pressure) the dominant loss from the plasma is synchrotron radiation, which has been neglected in obtaining the values of n, .tau. and T of the Lawson criterion mentioned above. Inclusion of this loss due to synchrotron radiation alters the Lawson Criterion to higher values of T and furthermore the fusion power output comes uncomfortably close to the synchrotron radiation output. Thus the margin for inefficiencies is much less. Detailed calculation by many authors has shown that the synchrotron radiation problem is minimal for a .beta. of at least 0.1. This value would also minimize the capital cost of a reactor. Thus values of about 20 times the best present tokamak values are needed. These values are partly related to the main method of heating the plasma, by driving a current of 100,000 A through it. Use of the neutral beam technology provides modest increases in .beta., perhaps factors of 2, but larger values seem unlikely. SUMMARY OF THE INVENTION It is known that the plasma parameters n, T and .tau. are determined mainly by the plasma current. However, the maximum value of such current I in amperes is limited by the Kruskal-Shafranov relation: ##EQU1## where R is the orbital radius and a is the radius of the cross section of the toroidal chamber. Thus, the attainable value of B.sub..phi. limits the maximum value of the plasma current I and consequently the plasma parameters. Instead of using higher toroidal fields B.sub..phi. with all their disadvantages to stabilize the plasma current this invention uses a magnetic field which is weaker than B.sub..phi.. This weaker magnetic field can be obtained from the strong focusing (s.f.) concepts elaborated by Courant et al. in Physical Review of 1952, at page 1190 and by U.S. Pat. No. 2,736,799 of N. Christofilos wherein the strong focusing was used for conventional particle accelerators such as synchrotrons. It should be noted that present tokamaks operate with a weak focusing (w.f.) field which is a cylindrically symmetric magnetic field at the orbit of the plasma current to keep the current channel fixed. This weak focusing field is quite weak, much weaker, for example, than the focusing produced by the magnetic field of the plasma current. This weak focusing magnetic field although sufficient to hold the plasma current channel in place, is not sufficient to stabilize against the Kruskal-Shafranov instability. A strong focusing magnetic field of strength 10 times larger than w.f. field should stabilize a plasma current channel with an aspect ratio R/a 5. Under these circumstances the s.f. force is about equal to the focusing of the current channel magnetic field (B.sub.s), for the circulating particles. Thus, the invention contemplates replacing the toroidal magnetic field of a tokamak with a strong focusing magnetic field. This strong focusing magnetic field is generally localized in the region of the toroidal chamber. To obtain the strong focusing the field itself is azimuthly non-uniform and in fact regularly varies in polarity for different azimuth positions. Such a fusion machine is hereinafter called a S. F. Megatron. |
Subsets and Splits
No community queries yet
The top public SQL queries from the community will appear here once available.