Split (2)

train · 33.5k rows

patent_number stringlengths 0 9	section stringclasses 4 values	raw_text stringlengths 0 954k
	abstract	A nuclear power plant spent fuel negative pressure unloading system comprises a fuel element transport pipe and a gas transport pipe. The fuel element transport pipe comprises a fuel element output pipe, a fuel element lifting pipe, and a fuel element unloading pipe connected in series. The fuel element unloading pipe is arranged obliquely downward in the direction of fuel element movement. The distal end of the fuel element unloading pipe is connected sequentially to fuel loading apparatus and a transfer apparatus. Two nozzles of the gas transport pipe are connected to set positions on the fuel element output pipe and the fuel element unloading pipe respectively. A gas driving mechanism is connected to the gas transport pipe. An inlet of the gas driving mechanism is arranged at one end in proximity to the fuel element unloading pipe.
039403092	summary	The present invention relates to a safety device for the control and safety rods of a nuclear reactor with vertical rods to the control drive mechanism that is to the mechanism for inserting into or withdrawing from the reactor the power control rods. This type of rods in addition to control functions, perform also a safety function in as much as the release thereof from the drive mechanism can be commanded which, in the case of vertical rods, results in the dropping of the rods into the reactor and as a consequence the shutdown of it. In fact those rods which are intended for performing the twofold task of control and safety rods are affixed to the drive mechanism through an electromagnetic coupling. This coupling comprises a portion which is attached to the drive mechanism and contains a core with the related energizing winding -- that is an electromagnet -- and a portion attached to the rod -- electromagnetic armature. While operating normally or even abnormally but within certain safe limits, the magnet is permanently energized and the rod is maintained suspended at a level which is determined by the drive mechanism. On the other hand, when particular conditions occur of the reactor operation, whereby the reactor must be shutdown, the power supply to the electromagnet is cut off and the rod drops into the reactor and causes that kind of shutdown which is called scram. Each of the two functions -- control and safety -- of the rod are dependent on a set of data different and independent from one of the others. In fact the data on which the operation of the drive mechanism is dependent are different by their own nature and by their magnitude from the data on which the release device is operated. Because this invention is intended for improving the means for achieving a safety control of the reactor with an increased safety factor, the drive mechanism and the related control means will be dealt with briefly. It should be noted that a failed shutdown of a reactor may bring about a disastrous situation, the untimely shutdown may also bring about undesirable consequences. From this the requirement ensues of providing the electromagnet coupling and the related controls with the highest possible reliability. The means which control the electromagnetic coupling include a number of sensing units or transducers sensitive to changes in various quantities such as the coolant temperature, the ratio of the reactor power and the coolant flow rate, the fuel temperature etc. The sensed data are fed to a series of combinational circuits which take into account the magnitude of the signals from the sensing units and evaluate all together to find out whether the conditions exist which impose the reactor shutdown. The input from the combinational circuit or logic lines are supplied to a feeder of the suspension electromagnet of the rod. In general in the existing plants, in order to improve the reliability of the safety control system, the number of sensing units of the logic lines of the feeders have been multiplied; however no such system is presently available wherein the magnet for suspending the rods is multiple. In other words, in the current plants, upstream of the electromagnet a high safety level is attained which is proportional to the number of logic lines connected in parallel. For example, of these logic lines, one can be assumed to be excluded due to a failure or for checking purposes without discontinuing the electromagnet supply. When determining the safety level of the lines by which the coupling is controlled, two aspects are to be considered, that is the requirement of ensuring the coupling effectiveness in the case also of a failure of a portion of said lines whereby it no longer supplies the coupling and the requirement of ensuring the release of the rod in the case of a failure of said portion of lines in a sense opposite to the former one, that is in the case that, although the conditions occur for the reactor shutdown, the safety rods still remain suspended because a portion of the coupling remains energized sufficiently to support the rod. In general the current safety systems are provided with three safety lines of which two at least must operate for obtaining the desired action. With this criterion, which hereinafter will be referred to as "two out of three logics" one of the three lines can suffer a failure or can be excluded for overhaul purposes without impairing the system effectiveness. However, with such systems, the servicing of a whole protective line to begin from the sensor of the reactor condition up to the electromagnet inclusive can be effected. This is a major limitation of the reliability level of the system. The main object of this invention is therefore to provide a suspension coupling for a control and safety rod which coupling comprises a plurality of electromagnets that have a plurality of electromagnet cores and related windings which are independent from one of the others. The magnetic fluxes of said plurality of electromagnets sum on a single armature which is attached to the rod. The total attractive force of the electromagnets is redundant with respect to the force required for ensuring a safe connection between the rod and the drive mechanism. Concurrently the redundant fraction of electromagnets should not be greater than that strictly required in as much as a failed de-energization of the redundant fraction when a release of the rod is required must not prevent such release. A coupling according to this invention permits the redundant fraction of electromagnets to be cut off for check and overhaul purposes without impairing the safety of the reactor.
	claims	1. Machinery system to cut out an approximately 20 feet diameter and approximately 36 inches thick steel-reinforced concrete circular segment at the top of a dome structure of a reactor primary containment, the system comprising,two diametrically opposite, movable cutting heads with adjustable cutting depth to gradually remove the concrete and steel around a perimeter at a certain radius of the circular segment of a planned dome opening,each cutting head equipped with multiple cutters with three axis directional movement and capable of rotating about its vertical axis at a set RPM, of vertically moving at a set rate and to a set cutting depth, and of traveling horizontally at a set feed speed in a circular arc forward from a starting point and reversing in an offset path from an end point back to the starting point in a repetitious manner. 2. The machinery system of claim 1, further comprising,a cross over structure for supporting the movable cutting heads, the structure capable of being temporarily anchored to the containment dome structure,a stationary circular inner and outer drive gear attached to the said cross over structure, andan inner and outer drive screw respectively engaged to the said drive gears and mounted on the movable cutting heads. 3. The machinery structure of claim 2, further comprising a lateral position setting and locking device mounted on a vertical support plate for reverse travel position of the cutting heads.
	abstract	In a method for measuring detected radiation, an analog data signal is converted to a digital data signal having aperiodic data pulses varying with intensity of the analog data signal. A time signal indicative of data intervals is produced. The data pulses are counted. A data count is stored in a start location and a corresponding time value is stored in a start location each time a data pulse occurs until a measured data interval starts. After a next data interval is detected, the data count is stored in an end location and a corresponding time value is stored in an end location when the next data pulse occurs. An average intensity of the detected radiation for the measured data interval is determined from the stored data counts and time values. A CT scanner (10) for measuring detected radiation includes a channel circuit (56), a storage circuit (60), a control circuit (58), and a processor (62).
	abstract	A supercritical pressure water cooled reactor comprising: a reactor vessel including: a shell part for containing sub-critical pressure coolant, and an end part for containing supercritical-pressure coolant which is separated from the sub-critical pressure coolant in the reactor vessel. A core-support plate with through-holes, the core-support plate disposed-in and fixed to the reactor vessel so that the core-support plate divides space inside the reactor vessel into a supercritical-pressure portion and a sub-critical pressure portion. Fuel tubes with both open ends fixed to the through-holes, the open ends being communicated to the supercritical-pressure portion, outside of the fuel tubes being disposed in the sub-critical pressure portion; and nuclear fuel assemblies disposed in the fuel tubes.
	abstract	A nuclear-fuel pin including a linear element made of a metal nuclear-fuel material consisting of uranium and/or plutonium, and cladding including Fe and Cr or an alloy including at least both of said elements, comprises a main shell provided around the linear nuclear-fuel element, said shell including threads or fibers made of SiC. A method for producing a nuclear-fuel pin is also provided.
	summary
	abstract	A method for developing a multi-focus primary collimator is described. The method includes extending a first beam from a first source via a collimator block to a first detector and determining a size of the first beam.
05999889&	summary	BACKGROUND OF THE INVENTION The present invention relates to the measurement of antenna performance parameters in real time. More specifically, but without limitation thereto, the present invention relates to the measurement of voltage, current, resistance, and reactance of a transmitting antenna during normal broadcast. Previous methods for measuring antenna performance typically required the transmission of a CW signal, which meant interrupting normal communications during the measurement period. The possibility of an undetected malfunction while the transmitter was operating is a factor limiting the confidence in the capability of a communications system. U.S. Pat. No. 5,233,537 issued on Aug. 3, 1993 and incorporated herein by reference thereto describes a system that solves the problem of testing an antenna while transmitting a frequency shift-keyed signal. FIG. 2 of this patent shows the four components used to measure antenna operating parameters: a sensor to generate a signal representative of antenna feed point voltage and current, a sampling trigger circuit, an A/D converter, and a data processing unit to calculate antenna performance from the sampled voltage and current. Because the antenna parameters change with the frequency of the signal being transmitted, the sampling trigger is required to gate samples to the A/D converter during the time periods when the frequency being measured is transmitted. Separate measurements are performed at each transmission frequency. A continuing need still exists, however, for an antenna performance monitor that is not dependent on knowledge of the frequency excursions of the transmitted signal. SUMMARY OF THE INVENTION An antenna performance monitor of the present invention is directed to overcoming the problems described above, and may provide further related advantages. No embodiment of the present invention described herein shall preclude other embodiments or advantages that may exist or become obvious to those skilled in the art. An antenna performance monitor of the present invention comprises an antenna sensor for coupling to an antenna radiating a radio frequency signal. The antenna sensor generates signals Vm and Im representative of the antenna input voltage and current. An A/D converter is coupled to the antenna sensor for digitizing signals Vm and Im. A data processor is coupled to the A/D converter for generating outputs representative of impedance magnitude and phase angle of the antenna substantially concurrently with changes in frequency of the radio frequency signal. An advantage of the antenna performance monitor is that the performance parameters of an antenna may be measured while connected to an operating transmitter. Another advantage is that no sampling trigger is required to select specific transmission frequencies. Yet another advantage is that antenna performance parameters may be broadcast virtually in real time within the transmitted signal or over other comunnication links. The features and advantages summarized above in addition to other aspects of the present invention will become more apparent from the description, presented in conjunction with the following drawings.
049873091	description	DESCRIPTION OF THE PREFERRED EMBODIMENTS Referring now to the drawings, wherein like reference numerals designate identical or corresponding parts, FIG. 1 shows a radiation therapy unit according to a preferred embodiment of the invention. It comprises a rotary stand 1, a radiator arm 2 mounted thereon and a radiator head 3. The patient to be treated lies on a treatment table 4 underneath the radiator head 3. The treatment table 4 is not a part of the application. It can be adjusted in height, rotated about a vertical axis and pushed forward and back in longitudinal and transverse direction, as is normal practice in the prior art. The rotary stand 1 is mounted rotatably about an axis A and is moved via a drive system 5. The components needed for generating high-energy electrons are accommodated in the rotary stand 1 and in the radiator arm 2. An RF source 7 controlled by a pulse transformer 6 generates microwaves of the power needed. The microwaves are fed via an RF feed 8 to an acceleration tube 9 which is accommodated in the radiator arm 2. An electron gun 10 injects electrons into the acceleration tube 9. The electrons are accelerated to, for example, 8MeV by the microwaves, filtered with respect to energy in a subsequent deflection magnet 11 and deflected in the direction of a beam axis S. The parts of the radiation therapy unit described until now are known as such. They can be replaced by other means which are suitable for generating a high-energy electron beam. The electrons emerging from the deflection magnet 11 can either be used themselves for the treatment or converted into photons by means of a target 12. The parts of the radiator head according to the invention develop their effect independently of whether photons or electrons or other particles are used in the therapy. This is why only a beam of rays is mentioned in the text following. The essential parts of the radiator head 3 are only indicated in FIG. 1. A target 12 (for generating the photons or an electron diffusing filter for diffusing the electrons) is located at the output of the deflection magnet 11 and generates the beam of rays. An important point of reference for the geometric-optical considerations is a point, hereinafter designated as focal point Q, in the three-dimensional space from which the beam of rays propagates in accordance with optical geometry. Another important point of reference in space is the isocenter I. This is the location at which the beam of rays develops the required optimum effect. Usually this is where the beam of rays encounters the tumor. Geometrically considered, the isocenter I is the point of intersection of the beam axis S with the axis of rotation A. The beam of rays which thus propagates along the beam axis S is transformed by a flattening filter 13 (for example into a beam of rays which is as homogeneous as possible over the maximum radiation field) and measured in an ionization chamber 14. Before the beam of rays leaves the radiator head 3, it is laterally limited by two pairs of collimators 17a, 17b which are aligned perpendicularly with respect to one another. If necessary, a wedge filter 15 can be pushed into the beam of rays underneath the ionization chamber 14. A mirror 16 and a light source 21 allow the lateral extent of the beam of rays to be made visible on the patient as simulation. According to a preferred embodiment, a matrix ionization chamber 19 is arranged on the beam axis S opposite to the radiator head 3 and below the isocenter. It is mounted at the rotary stand 1 by means of a fixed or retractable holder 18. FIG. 2 shows the radiator head 3. The target 12 defines the focal point Q on the beam axis S. Z designates a cone axis which extends perpendicular to the beam axis S through the focal point Q. A beam diaphragm 20 limits the radiation field to a circular area. The flattening filter 13 equalizes the radiation field (for example energy or dose peaks in the center). Instead of one single filter, several flattening filters can also be provided which are mounted on a slider. In the same manner, the target 12 can also be interchangeable. In the text which follows, the multi-leaf collimators according to the invention are explained. In this connection, the geometric reference points and axes defined above are of central significance. So that the description is not unnecessarily complicated, terms like "top" and "bottom" and "above" and "below" are used. They must be understood within the context that the focal point Q is "at the top" and the isocenter I is "at the bottom". Thus, the beam of rays propagates from top to bottom. Correspondingly, the formulation "object X is above object Y" means that object X is closer to the focal point than object Y. FIG. 2 shows an embodiment with two multi-leaf collimators which lie one above the other and are aligned perpendicularly with respect to one another and which in each case comprise two collimator halves 22a, 22b, 22c (the fourth collimator half is not drawn in favor of clear representation in FIG. 2). Each collimator half 22a, 22b, 22c is composed of a plurality of diaphragm plates 23 arranged next to one another which are in each case individually supported and guided by two holding yokes 24a (the second holding yoke of the collimator half of the top multi-leaf collimator is not visible in FIG. 2), 24b and 24c, 24d and 24e. The diaphragm plates 23 can be displaced independently of one another. In FIG. 2, for example, the diaphragm plates of the upper multi-leaf collimator move perpendicularly to the plane of the drawing and those of the lower multi-leaf collimator move in parallel with the plane of the drawing. For each diaphragm plate 23, a stepping motor 25 is provided which pushes this plate forward and back by means of a worm-rack gear 35 and a toothed rail 31. The stepping motors 25 of a multi-leaf collimator or of the corresponding collimator halves, respectively, are arranged to be staggered both vertically and horizontally. Naturally, instead of the stepping motor, any other drive (for example a direct-current motor with shaft encoder) can be used. The holding yokes guide, on the one hand, the diaphragm plates, and, on the other hand, support the worm-rack gears. In this arrangement, each diaphragm plate is supported and guided at four points by the two holding yokes of one collimator half. The advantage of the stepping motor lies in the fact that it is continuously under voltage and thus under control and that it always stops in a defined position. In consequence, it is possible to detect the position of the diaphragm plates via the stepping motors alone (without potentiometers). The worm-rack gear is self-inhibiting. This dispenses with an additional braking device which ensures that the diaphragm plates cannot shift independently. FIGS. 3a and 3b show a single diaphragm plate from the front and from the side, respectively. It has two side faces 26a, 26b, two front faces 27a, 27b, an inside face 28 and an outside face 29. The projection of the side faces 26a, 26b on a plane which is perpendicular to the cone axis (see later) has the shape of a circular ring segment. The diaphragm plate has the shape of a trapezoid with unequal sides in cross section, the parallel sides of the trapezoid being given by the inside and the outside face 28, 29 and the unequal sides of the trapezoid being given by the side faces 26a, 26b. The diaphragm plate has a height d and a thickness b. The inside face 28 is slightly narrower than the outside face 29 by exactly the amount that an extension of an edge between side face 26a and front face 27a and an extension of an edge between front face 27a and side face 26b (that is to say an extension of the legs of the trapezoid) intersect at the focal point Q (see FIG. 2). To minimize any leakage radiation between the diaphragm plates, the side faces 26a, 26b can be provided with mutually corresponding steps 30, 31. These would extend in parallel with the inside and outside face 28 and 29, respectively, that is to say they would be circular arc-shaped. FIG. 4 shows a collimator half having 20 diaphragm plates Each of these has at its outside face 29a, 29b, . . . a toothed rail 31a, 31b, . . . which is engaged by the previously mentioned worm-rack gear. The toothed rails 31a, 31b of adjacent diaphragm plates have a different height in each case. This makes it possible to accommodate the worm gears in the small space available. To obtain the same ratio between motor speed and plate advance for all diaphragm plates, the toothed rails also have a correspondingly different pitch. The two holding yokes of one collimator half are cut in in the shape of a comb on the inside of one external yoke. The toothed rails are inserted in the cut-ins and the appropriate diaphragm plate rests in each case with its outside face on adjacent teeth of the comb. Naturally, the diaphragm plates can also be supported and guided by other means. Thus, the outside faces can also be constructed, for example, to be V-shaped and mounted on rollers. Instead of attaching racks on the outside faces, adjusting means can be provided which directly engage the appropriately shaped outside face. In addition, beam axis S, focal point Q, cone axis Z and inside radius Ri and outside radius Ra are drawn in FIG. 4. The meaning of these geometric reference quantities is explained with reference to FIG. 5. FIG. 5 shows a spatial representation to explain the invention. The beam axis S and the cone axis Z intersect at a rightangle at the focal point Q. The cone axis Z is axis of an inner and of an outer cylinder having the radius Ri and Ra, respectively, Ri being<Ra. A plane perpendicular to the cone axis Z through the focal point Q is defined as center plane. Thus, the beam axis S is located in the center plane. The outer cylinder is now intersected with planes which are parallel to the center plane and equidistant from one another (distance b). The curves of intersection produced in this manner are circles. For each circle, an area is defined which is produced by the fact that the point of a vector originating from the focal point Q revolves on the circle. The area thus defined is in each case the envelope of a cone. Each cone envelope also intersects the inner cylinder. A space which is in each case bounded by two edges and cone envelopes and by the inner and the outer cylinder represents the path along which a diaphragm plate of the multi-leaf collimator according to the invention moves. The diaphragm plate itself is none other than a segment of this space bounded by angles. On the basis of this geometric consideration, the following becomes clear: 1. The multi-leaf collimator according to the invention is double-focussed. This is because in each position, the diaphragm plates are aligned both with the side and with the front faces in such a manner that a beam originating from the focal point can only be tangent to them but cannot intersect them. 2. The relationship between plate advance and speed of the motor is linear. This is because the outside faces of the diaphragm plates are all of equal length and move along the same cylinder surface. The outer and inner cylinder surface, respectively, is to be understood as enveloping surface. That is to say the outside and inside faces of the diaphragm plates do not need to be precisely a part of a cylinder surface. It is quite within the field of the invention to specially develop these faces for guiding purposes. FIG. 6 shows a three-dimensional representation of a diaphragm plate by means of which the features of the invention necessary in accordance with the concept of the invention are to be picked out. There are two of these: 1. Each side face 26a, 26b forms a part of a surface area of a cone K1, K2. All such cones K1, K2 have their point at the focal point Q. Furthermore, they have a common cone axis Z. It follows from this that the two cones K1, K2 belonging to one diaphragm plate have different slopes. On the other hand, the cones of side faces of adjacent diaphragm plates facing one another are identical. For this reason, adjacent diaphragm plates fit against one another in a form-closing manner. 2. The diaphragm plates must be guided by suitable means in such a manner that they execute a pure rotation about the common cone axis Z. The result is that the diaphragm plates do not move apart during the displacement. Thus, the side faces of adjacently arranged diaphragm plates also remain in form-closed contact. These two central points of the invention are unaffected by any other shape of the diaphragm plate (outside face, inside face, stepped side face and so forth) and the type of guiding means. FIG. 7 shows a representation of a radiation field such as can be generated by means of a radiator head having two multi-leaf collimators. A plane perpendicular to the beam axis through the isocenter is represented and a projection of the diaphragm plates on this plane. In the center of the figure, the area 36 to be irradiated is drawn. It has, for example, several indentations 36a, 36b, 36c which must be taken into account during the irradiation, in the sense that they should be protected against the radiation as far as possible. The diaphragm plates are then advanced by such an amount that an effective radiation area 33 approximates the area 36 to be irradiated as closely as possible. In FIG. 7, the shading indicates that 1. Certain areas (checked) are covered both by the upper and the lower multi-leaf collimator and PA1 2. Certain edge regions (obliquely shaded) are only covered by one of the two multi-leaf collimators. It has hitherto been necessary for each individual collimator to be at such a height that it was capable of completely attenuating the beam of rays, that is to say down to a required harmless level. If, however, two multi-leaf collimators according to the invention are used as in FIG. 2, the height d (see FIG. 3) of the diaphragm plates can be reduced. The reason for this is that each point of the radiation field can be selectively covered both by the upper and by the lower multi-leaf collimator. FIG. 8 illustrates the drive arrangement of multi-leaf collimators at reduced height. The same plane as in FIG. 7 is shown. The radiation field has a square maximum radiation area 32 of, for example, 40 cm.times.40 cm. The effect of radiation area 33 is understood to be the part of the radiation field which is not covered by diaphragm plates. In the present example, each of the four collimator halves is built up of 20 diaphragm plates. The diaphragm plates of the lower multi-leaf collimator move in the X direction in the representation of FIG. 7 and those of the upper multi-leaf collimator move in the Y direction (see coordinate system drawn) The thickness b of the individual diaphragm plates is selected in such a manner that their central projection from the focal point to a plane perpendicular to the beam axis S is of the same size. The diaphragm plates of the upper multi-leaf collimator are thus slightly narrower than those of the lower one. 23d designates a freely selected diaphragm plate of the lower and 23e one of the upper multi-leaf collimator. These two diaphragm plates together control a field point 34. Thus, if the two diaphragm plates 23d and 23e are advanced as shown in FIG. 7, it is sufficient if they are capable of covering the radiation field together. Thus, the height d can be reduced compared with a conventional collimator. In principle, the height can be reduced to one half. However, the height d is advantageously only reduced by about one third compared with the original value. This is because such a height already produces a reduction in the intensity to a few percent. In the case of a region to be irradiated such as, for example, in FIG. 7, the indentations 36a, 36b, 36c to be protected can in most cases already be protected sufficiently well against radiation damage even though they are only covered by the diaphragm plates of one multi-leaf collimator. (The obliquely shaded regions which, of course, cannot be covered at all by means of conventional collimator blocks are thus only loaded with a few per cent of the radiation dose). A radiation therapy unit having multi-leaf collimators with reduced height must comprise a control circuit which controls the diaphragm plates in such a manner that each field point to be covered is always covered both by a diaphragm plate of the upper and by one of the lower multi-leaf collimators. Such a control circuit can be implemented best by programmed microprocessor. The advantage of this embodiment lies in the fact that there is more space available in the radiator head or, conversely, the constructional height becomes smaller and that the weight of the multi-leaf collimators becomes less. A preferred embodiment of the invention comprises a matrix ionization chamber 19 for monitoring the radiation field. Radiation therapy units are subject to certain safety regulations. This also includes, for example, the regulation that the collimators must be monitored by two separate channels. One of these channels is the stepping motor which holds the diaphragm plate in an accurately defined position. The other channel is now the matrix ionization chamber 19 (see FIG. 1). It is arranged below the isocenter I on the beam axis S. Thus, it detects the shape of the radiation field in the way in which it acts on the patient, and that almost in real time. If a fault occurs in the multi-leaf collimator, the therapy can be immediately interrupted. This arrangement also makes it possible to change the radiation field during the therapy. This is of advantage, for example, when the direction of irradiation is changed during the treatment (rotation around the isocenter I). The matrix ionization chamber as such is known. This is why it will not be discussed in greater detail at this point but express reference is made to the published European patent application EP-0,196,138 A2. The invention can be realized in the most varied manners. In the text which follows, a few further possibilities will be indicated briefly. Naturally, potentiometers can also be used for monitoring the diaphragm plates. In particular, these can handle the monitoring function of the matrix ionization chamber as an alternative. The reduction of the leakage radiation between the diaphragm plates can also be achieved by means of slot-like recesses instead of by means of the steps described. In this connection, it is understood that the recesses extend in parallel with the inside and outside face and allow an essentially form-closed contact between adjacent diaphragm plates. To keep the number of actuators needed for driving the multi-leaf collimators as small as possible, the stepping motors are advantageously operated in multiplex mode. The diaphragm plates consist of a material conventionally used for collimators. Tungsten alloys should be mentioned as an example. If then a height d=70 mm of such a material is required for covering the radiation field, a height of d=35 mm is sufficient for a multi-leaf collimator in the embodiment with reduced height. A preferred height would be about 50 mm. Finally, it can be said that the invention creates a radiation therapy unit which can be used in many different ways. Obviously, numerous modifications and variations of the present invention are possible in light of the above teachings. It is therefore to be understood that within the scope of the appended claims, the invention may be practised otherwise than as specifically described herein.
041644794	description	DESCRIPTION OF THE PREFERRED EMBODIMENT These and other objects of the invention for an improvement in the method of suppressing fluoride and chloride volatility during the calcination of aqueous nuclear fuel reprocessing solutions containing zirconium fluoride, chloride and other values by adding calcium nitrate to the solution in an amount sufficient to establish a calcium to fluoride mole ratio of at least 0.55 wherein the calcium nitrate acts to suppress fluoride volatility during calcination, the improvement wherein aluminum nitrate is added to the solution before the calcium nitrate in an amount sufficient to make the aluminum to fluoride mole ratio from about 0.27 to 0.40, whereby the formation of gelatinous solids in the waste due to the presence of calcium nitrate is substantially reduced and the volatility of the chloride during calcination of the solution is suppressed. This invention is particularly suited for suppressing fluoride and chloride volatility while reducing the gelatinous solids formed by the addition of calcium nitrate to aqueous waste solutions such as the first cycle zirconium-fluoride waste resulting from the reprocessing of zirconium fuels at the Idaho Chemical Processing Plant (ICPP) and to the blend of first cycle waste with second cycle waste from zirconium fuel reprocessing. It is also suitable for improving the calcinability of any aqueous solution containing zirconium, fluoride and chloride compounds. In Table I below are given the nominal compositions of the two waste solutions. TABLE I ______________________________________ COMPOSITION OF WASTES Concentrations 1st Cycle Zirconium- 2nd Cycle Constituent Fluoride Waste Waste ______________________________________ H.sup.+ 2.3 M 1.3 M Zr 0.38 M Al 0.64 M 0.66 M Cr 1.6 g/l Sn 0.39 g/l B 1.7 g/l 0.14 g/l Na 59.5 g/l K 9.6 g/l Fe 1.2 g/l Mn 1350 ppm NH 0.035 M Hg.sup.4 0.99 g/l NO.sub.3.sup.- 2.8 M 6.0 M F.sup.- 3.0 M 0.0065 M PO.sub.4.sup.-3 2.7 g/l SO.sub.4.sup.-2 0.063 M Cl trace 1700 ppm ______________________________________ Calcium nitrate is added to the first cycle zirconium-fluoride waste solution in an amount sufficient to make the calcium to fluoride mole ratio at least 0.55 to provide adequate suppression of the fluoride volatility. Although this is sufficient for first cycle waste, a mole ratio of at least 0.6, preferably 0.7, is necessary when the blend of wastes is calcined. This is required to prevent nodules forming on the fluidized-bed material and ultimately causing a collapsed bed. These nodules are believed to be due to sodium in the second cycle waste. It will be noted in Table I, that both waste solutions contain aluminum, the first cycle waste having a normal aluminum to fluoride mole ratio of about 0.21 while the blend has a ratio of about 0.28. The amount of aluminum to be added to first cycle waste must be an amount sufficient to establish a mole ratio of aluminum to fluoride from about 0.27 to about 0.40. Although the 0.27 ratio is preferred, increased aluminum content was found to have no deleterious effects. Although the blend of first cycle and second cycle wastes, contains sufficient aluminum to establish an aluminum to fluoride mole ratio of 0.28, this is insufficient to provide adequate chloride volatility suppression for reasons unknown. However, when sufficient aluminum is added to establish an aluminum to fluoride mole ratio from about 0.32 to about 0.4, with 0.32 being preferred, the volatility of the chloride present in the blend was substantially reduced. The calcium and aluminum are generally added to the waste solutions as nitrates because of solubility and compatibility with the compounds already present, although any compound, which is soluble in the solution and compatible with the ions already present, would be suitable. The reasons for the effect of the increased aluminum to fluoride ratio on reducing the amount of gelatinous solids formed by the addition of calcium nitrate and on the suppression of chloride volatility are unknown. The following examples are given to show the operability of the method of the invention and are not to be taken as limiting the scope of the invention as defined by the claims appended hereto. EXAMPLE I To demonstrate the effect of the addition of calcium and aluminum on the amount of gelatinous solids formed in the first cycle waste and in the blend, experiments were run in which varying amounts of ions were added to the wastes. In Table II the rate of filtration of solids after calcium nitrate or aluminum nitrate plus calcium nitrate had been added to the wastes is used as a measure of the gelatinous nature of the residue -- the less the filtering time, the less the gelatinous nature of the solid. In each case 30 ml of homogenized slurry is sucked through a sintered glass filter (that has never been used before) having a 14 micron porosity by a vacuum pressure of 17 inches of mercury. The results are given in Table II below. TABLE II ______________________________________ Effect of Calcium and Aluminum Concentrations On The -Gelatinous Nature And Amount Of Solids Formed In Zirconium And Fluoride-Containing Wastes Residue Ca to F Al to F (g from Mole Mole Filtering 30 ml Waste Ratio Ratio Time waste) ______________________________________ 1st Cycle 0.55 0.21 25 min 4.6 Zr-F Waste 1st Cycle 0.55 0.27 5 min 2.6 Zr-F Waste 10 sec 1st Cycle 0.55 0.40 1 min 0.55 Zr-F Waste 5 sec 3 Vol 1st 0.7 0.28 3 min 2.4 Cycle Zr-F Waste blended with 1 vol 2nd Cycle Waste 3 Vol 1st 0.7 0.32 45 sec 0.52 Cycle Zr-F Waste Blend with 1 vol 2nd Cycle Waste ______________________________________ It can be seen that the addition of a small amount of aluminum resulted in a substantial reduction of the amount of solids formed. EXAMPLE II The method for decreasing gelatinous solids in calciner feed was tested by runs in a 4-inch diameter, fluidized-bed, in-bed combustion, pilot plant calciner to determine how the methods affected fluoride and chloride volatility, calciner operability, and calcine properties in such a calciner. Table III shows that increasing the aluminum to fluoride mole ratio in first-cycle zirconium-fluoride waste from 0.21 to 0.40 prior to Ca(NO.sub.3).sub.2 addition had no adverse effect on fluoride volatility, calciner operability, and calcine properties. The attrition index is a measure of the hardness of bed particles -- the smaller the index, the softer the particles. Table III also shows that the use of Mg(NO.sub.3).sub.2 produces a bed particle that is probably too soft; a soft bed particle breaks easily into fines during fluidized-bed operation, and the production of too many fines would likely result in plugging and bridging in the calciner off-gas and transport systems. Thus, the use of Mg(NO.sub.3).sub.2 is not recommended for use in the calcination of first-cycle zirconium-fluoride waste. Table IV shows that increasing the aluminum to fluoride mole ratio from 0.28 to 0.32 in a blend of three volumes of first-cycle zirconium fluoride waste with one volume of second-cycle waste prior to calcium nitrate addition reduced chloride volatility, suppressed fluoride volatility satisfactorily, resulted in smooth calciner operation and produced a calcine of acceptable properties. TABLE III ______________________________________ Calcination Of First-Cycle Zirconium-Fluoride Waste In A 4-Inch Diameter, Fluidized-Bed, In-Bed -Combustion Calciner Run # FV4-lb FV4-2 FV4-3 ______________________________________ Run Duration (Hrs.) 58.7 40 40 Ca/F Mole Ratio 0.55 0.55 0 Mg/f Mole Ratio 0 0 0.55 Al/F Mole Ratio 0.21 0.40 0.21 Wt % Volatilized 0.6 0.1 0.2 from Calciner Calcination Temp. (.degree.C.) 500 500 500 Product to Fines 2.76 2.01 1.66 Ratio Density of Product 1.22 1.21 1.15 (g/cc) Density of Fines 0.54 0.77 0.57 11 (g/cc) Attrition Index of 28 16 4 the Final Bed (of the -32 +35 Mesh Fraction) (%) Calciner Operability No No No problems problems problems ______________________________________ TABLE IV ______________________________________ Calcination Of A Blend Of 3 Volumes First-Cycle Zirconium-Fluoride Waste With 1 Volume Second-Cycle Waste In Fluidized-Bed, In-Bed Combustion Calciner Run # 53 FV4-4 SBW 4-9 ______________________________________ Run Duration (Hrs.) 131 72 40 Calcination Temp. (..degree.C.) 500 500 500 Ca/F Mole Ratio 0.7 0.7 0.7 Al/F Mole Ratio 0.28 0.32 0.32 Wt % F Volatilized (a) 0.2 0.7 from Calciner Wt % Cl Retained in 70 92 92 Bed Plus Fines Product to Fines (a) 2.77 5.6 Ratio Density of Product (a) 1.58 1.68 (g/cc) Density of Fines 0.49 0.46 0.65 (g/cc) Attrition Index of 68 76 80 the Final Bed (of the -32 +35 Mesh Fraction) (%) Calciner Operability No No No problems problems problems Texture of Smooth Smooth Smooth calcine surface ______________________________________ Run 53 was made in a 12 inch diameter fluidized bed, in-bed combustion calciner Run FV4-4, SBW 4-9 - were made 4 inch diameter fluidized bed, in bed combustion calciner.? As can be seen from the preceding discussion and Examples, the addition of aluminum to the zirconium-fluoride waste and in the blend prior to adding calcium nitrate, not only reduces the volume of gelatinous solids formed by the calcium nitrate but also substantially decreases the volatility of the chloride in the blend. Reducing chloride volatility helps not only to reduce equipment corrosion but also reduces the possibility of corrosive gases escaping into and polluting the environment. Reducing volume of gelatinous solids reduces the potential for plugging the feed system to a calciner.
	claims	1. A system for installing and operating a control rod blade in a Nordic reactor pressure vessel (RPV) of a nuclear powerplant, the system comprising:a control rod blade (CRB), wherein the CRB extends in an axial direction and includes a neutron absorber section and a single-piece lower fin joined directly below the neutron absorber section, wherein the lower fin includes a bottom terminal edge with a receptor, and wherein the entire control rod blade is shaped and sized to fit between a plurality of fuel bundles in the RPV;a control rod drive system configured for linearly moving the CRB in the axial direction; andan adapter having a single-piece construction, wherein,the single-piece adapter includes a top axial end shaped to directly mate with the receptor,the single-piece adapter includes a bottom axial end with a fastener that removably connects directly to the control rod drive,the single-piece adapter has an axial length spanning an entire distance between the receptor in the bottom terminal edge of the CRB and at least the top of the control rod drive, andthe single-piece adapter is configured to connect the single-piece lower fin of the CRB and the control rod drive system via only the single-piece adapter. 2. The system of claim 1, wherein the CRB is cruciform and is configured to fit between the fuel bundles in to perpendicular transverse directions. 3. The system of claim 1, wherein the top axial end of the single-piece adapter is configured to be welded with the receptor in the bottom terminal edge of the control rod blade, and wherein the adapter supports a weight of the control rod blade as the control rod blade is moved in a linear motion. 4. The system of claim 3, wherein the single-piece adapter is cylindrical. 5. The system of claim 4, wherein the axial length of the single-piece adapter is about 95 inches to about 105 inches. 6. The system of claim 5, wherein the single-piece adapter has an outside cylindrical diameter of about 2.5 inches to about 3.5 inches. 7. A method of using the system of claim 1, the method comprising:lowering the CRB through the fuel bundles so that the receptor of the bottom terminal edge of the single-piece lower fin of the CRB is accessible below the fuel bundles;connecting the single-piece adaptor to the CRB by mating the top axial end of the single-piece adaptor directly to the receptor of the bottom terminal edge of the single-piece lower fin of the CRB; andremovably attaching the bottom axial end of the single-piece adaptor directly to the control rod drive system. 8. The method of claim 7, wherein the mating includes welding the top axial end of the single-piece adaptor directly to the receptor of the bottom terminal edge of the single-piece lower fin of the CRB. 9. The method of claim 7, wherein the removably attaching includes fastening the bottom axial end of the single-piece adaptor directly to the control rod drive system. 10. The method of claim 7, further comprising:driving the control rod drive system to move the CRB in the axial direction.
	description	The present application is a U.S. national stage application under 35 U.S.C. 371 of PCI Application No. PCT/US2013/042070 filed May 21, 2013, which claims the benefit of U.S. Provisional Patent Application Ser. No. 61/649,593 filed May 21, 2012, the entireties of which are incorporated herein by reference. The present invention relates nuclear reactors, and more particularly to a reactor containment system with passive thermal energy release control. The containment for a nuclear reactor is defined as the enclosure that provides environmental isolation to the nuclear steam supply system (NSSS) of the plant in which nuclear fission is harnessed to produce pressurized steam. A commercial nuclear reactor is required to be enclosed in a pressure retaining structure which can withstand the temperature and pressure resulting from the most severe accident that can be postulated for the facility. The most severe energy release accidents that can be postulated for a reactor and its containment can be of two types. First, an event that follows a loss-of-coolant accident (LOCA) and involve a rapid large release of thermal energy from the plant's nuclear steam supply system (NSSS) due to a sudden release of reactor's coolant in the containment space. The reactor coolant, suddenly depressurized, would violently flash resulting in a rapid rise of pressure and temperature in the containment space. The in-containment space is rendered into a mixture of air and steam. LOCA can be credibly postulated by assuming a sudden failure in a pipe carrying the reactor coolant. Another second thermal event of potential risk to the integrity of the containment is the scenario wherein all heat rejection paths from the plant's nuclear steam supply system (NSSS) are lost, forcing the reactor into a “scram.” A station black-out is such an event. The decay heat generated in the reactor must be removed to protect it from an uncontrolled pressure rise. More recently, the containment structure has also been called upon by the regulators to withstand the impact from a crashing aircraft. Containment structures have typically been built as massive reinforced concrete domes to withstand the internal pressure from LOCA. Although its thick concrete wall could be capable of withstanding an aircraft impact, it is also a good insulator of heat, requiring pumped heat rejection systems (employ heat exchangers and pumps) to reject its unwanted heat to the external environment (to minimize the pressure rise or to remove decay heat). Such heat rejection systems, however, rely on a robust power source (off-site or local diesel generator, for example) to power the pumps. The station black out at Fukushima in the wake of the tsunami is a sobering reminder of the folly of relying on pumps. Present day containment structures with their monolithic reinforced concrete construction make it extremely difficult and expensive to remove and install a large capital requirement such as a steam generator in the NSSS enclosed by them. To make a major equipment change out, a hatch opening in the thick concrete dome has to be made at great expense and down time for the reactor. Unfortunately, far too many steam generators have had to be changed out at numerous reactors in the past 25 years by cutting through the containment dome at billions of dollars in cost to the nuclear power industry. The above weaknesses in the state-of-the-art call for an improved nuclear reactor containment system. The present invention provides nuclear reactor containment system that overcomes the deficiencies of the foregoing arrangements. The containment system generally includes an inner containment vessel which may be formed of steel or another ductile material and an outer containment enclosure structure (CES) thereby forming a double walled containment system. In one embodiment, a water-filled annulus may be provided between the containment vessel and the containment enclosure structure providing an annular cooling reservoir. The containment vessel may include a plurality of longitudinal heat transfer fins which extend (substantially) radial outwards from the vessel in the manner of “fin”. The containment vessel thus serves not only as the primary structural containment for the reactor, but is configured and operable to function as a heat exchanger with the annular water reservoir acting as the heat sink. Accordingly, as further described herein, the containment vessel advantageously provides a passive (i.e. non-pumped) heat rejection system when needed during a thermal energy release accident such as a LOCA or reactor scram to dissipate heat and cool the reactor. In one embodiment according to the present disclosure, a nuclear reactor containment system includes a containment vessel configured for housing a nuclear reactor, a containment enclosure structure (CES) surrounding the containment vessel, and an annular reservoir formed between the containment vessel and containment enclosure structure (CES) for extracting heat energy from the containment space. In the event of a thermal energy release incident inside the containment vessel, heat generated by the containment vessel is transferred to the annular reservoir which operates to cool the containment vessel. In one embodiment, the annular reservoir contains water for cooling the containment vessel. A portion of the containment vessel may include substantially radial heat transfer fins disposed in the annular reservoir and extending between the containment vessel and containment enclosure structure (CES) to improve the dissipation of heat to the water-filled annular reservoir. When a thermal energy release incident occurs inside the containment vessel, a portion of the water in the annulus is evaporated and vented to atmosphere through the containment enclosure structure (CES) annular reservoir in the form of water vapor. Embodiments of the system may further include an auxiliary air cooling system including a plurality of vertical inlet air conduits spaced circumferentially around the containment vessel in the annular reservoir. The air conduits are in fluid communication with the annular reservoir and outside ambient air external to the containment enclosure structure (CES). When a thermal energy release incident occurs inside the containment vessel and water in the annular reservoir is substantially depleted by evaporation, the air cooling system becomes operable by providing a ventilation path from the reservoir space to the external ambient. The ventilation system can thus be viewed as a secondary system that can continue to cool the containment ad infinitum. According to another embodiment, a nuclear reactor containment system includes a containment vessel configured for housing a nuclear reactor, a containment enclosure structure (CES) surrounding the containment vessel, a water filled annulus formed between the containment vessel and containment enclosure structure (CES) for cooling the containment vessel, and a plurality of substantially radial fins protruding outwards from the containment vessel and located in the annulus. In the event of a thermal energy release incident inside the containment vessel, heat generated by the containment vessel is transferred to the water filled reservoir in the annulus through direct contact with the external surface of the containment vessel and its fins substantially radial thus cooling the containment vessel. In one embodiment, when a thermal energy release incident occurs inside the containment vessel and water in the annulus is substantially depleted by evaporation, the air cooling system is operable to draw outside ambient air into the annulus through the air conduits to cool the heat generated in the containment (which decreases exponentially with time) by natural convection. The existence of water in the annular region completely surrounding the containment vessel will maintain a consistent temperature distribution in the containment vessel to prevent warping of the containment vessel during the thermal energy release incident or accident. In another embodiment, a nuclear reactor containment system includes a containment vessel including a cylindrical shell configured for housing a nuclear reactor, a containment enclosure structure (CES) surrounding the containment vessel, an annular reservoir containing water formed between the shell of the containment vessel and containment enclosure structure (CES) for cooling the containment vessel, a plurality of external (substantially) radial fins protruding outwards from the containment vessel into the annulus, and an air cooling system including a plurality of vertical inlet air conduits spaced circumferentially around the containment vessel in the annular reservoir. The air conduits are in fluid communication with the annular reservoir and outside ambient air external to the containment enclosure structure (CES). In the event of a thermal energy release incident inside the containment vessel, heat generated by the containment vessel is transferred to the annular reservoir via the (substantially) radial containment wall along with its internal and external fins which operates to cool the containment vessel. Advantages and aspects of a nuclear reactor containment system according to the present disclosure include the following: Containment structures and systems configured so that a severe energy release event as described above can be contained passively (e.g. without relying on active components such as pumps, valves, heat exchangers and motors); Containment structures and systems that continue to work autonomously for an unlimited duration (e.g. no time limit for human intervention); Containment structures fortified with internal and external ribs (fins)configured to withstand a projectile impact such as a crashing aircraft without losing its primary function (i.e. pressure & radionuclide (if any) retention and heat rejection); and Containment vessel equipped with provisions that allow for the ready removal (or installation) of major equipment through the containment structure. All drawings are schematic and not necessarily to scale. The features and benefits of the invention are illustrated and described herein by reference to illustrative embodiments. This description of illustrative embodiments is intended to be read in connection with the accompanying drawings, which are to be considered part of the entire written description. In the description of embodiments disclosed herein, any reference to direction or orientation is merely intended for convenience of description and is not intended in any way to limit the scope of the present invention. Relative terms such as “lower,” “upper,” “horizontal,” “vertical,”, “above,” “below,” “up,” “down,” “top” and “bottom” as well as derivative thereof (e.g., “horizontally,” “downwardly,” “upwardly,” etc.) should be construed to refer to the nominal orientation as then described or as shown in the drawing under discussion. These relative terms are for convenience of description only and do not require that the apparatus be constructed or operated in a rigorously specific orientation denoted by the term. Terms such as “attached,” “affixed.” “connected.” “coupled,” “interconnected,” and similar refer to a relationship wherein structures are secured or attached to one another either directly or indirectly through intervening structures, as well as both movable or rigid attachments or relationships, unless expressly described otherwise. Accordingly, the disclosure expressly should not be limited to such illustrative embodiments illustrating some possible non-limiting combination of features that may exist alone or in other combinations of features. Referring to FIGS. 1-15, a nuclear reactor containment system 100 according to the present disclosure is shown. The system 100 generally includes an inner containment structure such as containment vessel 200 and an outer containment enclosure structure (CES) 300 collectively defining a containment vessel-enclosure assembly 200-300. The containment vessel 200 and containment enclosure structure (CES) 300 are vertically elongated and oriented, and defines a vertical axis VA. In one embodiment, the containment vessel-enclosure assembly 200-300 is configured to be buried in the subgrade at least partially below grade (see also FIGS. 6-8). The containment vessel-enclosure assembly 200-300 may be supported by a concrete foundation 301 comprised of a bottom slab 302 and vertically extending sidewalls 303 rising from the slab forming a top base mat 304. The sidewalls 303 may circumferentially enclose containment vessel 200 as shown wherein a lower portion of the containment vessel may be positioned inside the sidewalls. In some embodiments, the sidewalls 303 may be poured after placement of the containment vessel 200 on the bottom slab 302 (which may be poured and set first) thereby completely embedding the lower portion of the containment vessel 200 within the foundation. The foundation walls 303 may terminate below grade in some embodiments as shown to provide additional protection for the containment vessel-enclosure assembly 200-300 from projectile impacts (e.g. crashing plane, etc.). The foundation 301 may have any suitable configuration in top plan view, including without limitation polygonal (e.g. rectangular, hexagon, circular, etc.). In one embodiment, the weight of the containment vessel 200 may be primarily supported by the bottom slab 302 on which the containment vessel rests and the containment enclosure structure (CES) 300 may be supported by the base mat 304 formed atop the sidewalls 303 of the foundation 301. Other suitable vessel and containment enclosure structure (CES) support arrangements may be used. With continuing reference to FIGS. 1-15, the containment structure 200 may be an elongated vessel 202 including a hollow cylindrical shell 204 with circular transverse cross-section defining an outer diameter D1, a top head 206, and a bottom head 208. In one embodiment, the containment vessel 200 (i.e. shell and heads) may be made from a suitably strong and ductile metallic plate and bar stock that is readily weldable (e.g. low carbon steel). In one embodiment, a low carbon steel shell 204 may have a thickness of at least 1 inch. Other suitable metallic materials including various alloys may be used. The top head 206 may be attached to the shell 204 via a flanged joint 210 comprised of a first annular flange 212 disposed on the lower end or bottom of the top head and a second mating annular flange 214 disposed on the upper end or top of the shell. The flanged joint 210 may be a bolted joint, which optionally may further be seal welded after assembly with a circumferentially extending annular seal weld being made between the adjoining flanges 212 and 214. The top head 206 of containment vessel 200 may be an ASME (American Society of Mechanical Engineers) dome-shaped flanged and dished head to add structural strength (i.e. internal pressure retention and external impact resistance), however, other possible configurations including a flat top head might be used. The bottom head 208 may similarly be a dome-shaped dished head or alternatively flat in other possible embodiments. In one containment vessel construction, the bottom head 208 may be directly welded to the lower portion or end of the shell 204 via an integral straight flange (SF) portion of the head matching the diameter of shell. In one embodiment, the bottom of the containment vessel 200 may include a ribbed support stand 208a or similar structure attached to the bottom head 208 to help stabilize and provide level support for the containment vessel on the slab 302 of the foundation 301, as further described herein. In some embodiments, the top portion 216 of the containment vessel shell 204 may be a diametrically enlarged segment of the shell that forms a housing to support and accommodate a polar crane (not shown) for moving equipment, fuel, etc. inside the containment vessel. This will provide crane access to the very inside periphery of the containment vessel and enable placement of equipment very close to the periphery of the containment vessel 200 making the containment vessel structure compact. In one configuration, therefore, the above grade portion of the containment vessel 200 may resemble a mushroom-shaped structure. In one possible embodiment, the enlarged top portion 216 of containment vessel 200 may have an outer diameter D2 that is larger than the outer diameter D1 of the rest of the adjoining lower portion 218 of the containment vessel shell 204. In one non-limiting example, the top portion 216 may have a diameter D2 that is approximately 10 feet larger than the diameter D1 of the lower portion 218 of the shell 204. The top portion 216 of shell 204 may have a suitable height H2 selected to accommodate the polar crane with allowance for working clearances which may be less than 50% of the total height H1 of the containment vessel 200. In one non-limiting example, approximately the top ten feet of the containment vessel 200 (H2) may be formed by the enlarged diameter top portion 216 in comparison to a total height H1 of 200 feet of the containment vessel. The top portion 216 of containment vessel 200 may terminate at the upper end with flange 214 at the flanged connection to the top head 206 of the containment vessel. In one embodiment, the diametrically enlarged top portion 216 of containment vessel 200 has a diameter D2 which is smaller than the inside diameter D3 of the containment enclosure structure (CES) 300 to provide a (substantially) radial gap or secondary annulus 330 (see, e.g. FIG. 4). This provides a cushion of space or buffer region between the containment enclosure structure (CES) 300 and containment vessel top portion 216 in the advent of a projectile impact on the containment enclosure structure (CES). Furthermore, the annulus 330 further significantly creates a flow path between primary annulus 313 (between the shells of the containment enclosure structure (CES) 300 and containment vessel 200) and the head space 318 between the containment enclosure structure (CES) dome 316 and top head 206 of the containment vessel 200 for steam and/or air to be vented from the containment enclosure structure (CES) as further described herein. Accordingly, the secondary annulus 330 is in fluid communication with the primary annulus 313 and the head space 318 which in turn is in fluid communication with vent 317 which penetrates the dome 316. In one embodiment, the secondary annulus 330 has a smaller (substantially) radial width than the primary annulus 313. Referring to FIGS. 1-4, the containment enclosure structure (CES) structure (CES) 300 may be double-walled structure in some embodiments having sidewalls 320 formed by two (substantially) radially spaced and interconnected concentric shells 310 (inner) and 311 (outer) with plain or reinforced concrete 312 installed in the annular space between them. The concentric shells 310, 311 may be made of any suitably strong material, such as for example without limitation ductile metallic plates that are readily weldable (e.g. low carbon steel). Other suitable metallic materials including various alloys may be used. In one embodiment, without limitation, the double-walled containment enclosure structure (CES) 300 may have a concrete 312 thickness of 6 feet or more which ensures adequate ability to withstand high energy projectile impacts such as that from an airliner. The containment enclosure structure (CES) 300 circumscribes the containment vessel shell 204 and is spaced (substantially) radially apart from shell 204, thereby creating primary annulus 313. Annulus 313 may be a water-filled in one embodiment to create a heat sink for receiving and dissipating heat from the containment vessel 200 in the case of a thermal energy release incident inside the containment vessel. This water-filled annular reservoir preferably extends circumferentially for a full 360 degrees in one embodiment around the perimeter of upper portions of the containment vessel shell 204 lying above the concrete foundation 301. FIG. 4 shows a cross-section of the water-filled annulus 313 without the external (substantially) radial fins 221 in this figure for clarity. In one embodiment, the annulus 313 is filled with water from the base mat 304 at the bottom end 314 to approximately the top end 315 of the concentric shells 310, 311 of the containment enclosure structure (CES) 300 to form an annular cooling water reservoir between the containment vessel shell 204 and inner shell 310 of the containment enclosure structure (CES). This annular reservoir may be coated or lined in some embodiments with a suitable corrosion resistant material such as aluminum, stainless steel, or a suitable preservative for corrosion protection. In one representative example, without limitation, the annulus 313 may be about 10 feet wide and about 100 feet high. In one embodiment, the containment enclosure structure (CES) 300 includes a steel dome 316 that is suitably thick and reinforced to harden it against crashing aircraft and other incident projectiles. The dome 316 may be removably fastened to the shells 310, 311 by a robust flanged joint 318. In one embodiment, the containment enclosure structure (CES) 300 is entirely surrounded on all exposed above grade portions by the containment enclosure structure (CES) 300, which preferably is sufficiently tall to provide protection for the containment vessel against aircraft hazard or comparable projectile to preserve the structural integrity of the water mass in the annulus 313 surrounding the containment vessel. In one embodiment, as shown, the containment enclosure structure (CES) 300 extends vertically below grade to a substantial portion of the distance to the top of the base mat 304. The containment enclosure structure (CES) 300 may further include at least one rain-protected vent 317 which is in fluid communication with the head space 318 beneath the dome 316 and water-filled annulus 313 to allow water vapor to flow, escape, and vent to atmosphere. In one embodiment, the vent 317 may be located at the center of the dome 316. In other embodiments, a plurality of vents may be provided spaced (substantially) radially around the dome 316. The vent 317 may be formed by a short section of piping in some embodiments which is covered by a rain hood of any suitable configuration that allows steam to escape from the containment enclosure structure (CES) but minimizes the ingress of water. In some possible embodiments, the head space 318 between the dome 316 and top head 206 of the containment vessel 200 may be filled with an energy absorbing material or structure to minimize the impact load on the containment enclosure structure (CES) dome 316 from a crashing (falling) projecting (e.g. airliner, etc.). In one example, a plurality of tightly-packed undulating or corrugated deformable aluminum plates may be disposed in part or all of the head space to form a crumple zone which will help absorb and dissipate the impact forces on the dome 316. Referring primarily to FIGS. 1-5 and 8-17, the buried portions of the containment vessel 200 within the concrete foundation 301 below the base mat 304 may have a plain shell 204 without external features. Portions of the containment vessel shell 204 above the base mat 304, however, may include a plurality of longitudinal external (substantially) radial ribs or fins 220 which extend axially (substantially) parallel to vertical axis VA of the containment vessel-enclosure assembly 200-300. The external longitudinal fins 220 are spaced circumferentially around the perimeter of the containment vessel shell 204 and extend (substantially) radially outwards from the containment vessel. The ribs 220 serve multiple advantageous functions including without limitation (1) to stiffen the containment vessel shell 204, (2) prevent excessive “sloshing” of water reserve in annulus 313 in the occurrence of a seismic event, and (3) significantly to act as heat transfer “fins” to dissipate heat absorbed by conduction through the shell 204 to the environment of the annulus 313 in the situation of a fluid/steam release event in the containment vessel. Accordingly, in one embodiment to maximize the heat transfer effectiveness, the longitudinal fins 220 extend vertically for substantially the entire height of the water-filled annulus 313 covering the effective heat transfer surfaces of the containment vessel 200 (i.e. portions not buried in concrete foundation) to transfer heat from the containment vessel 200 to the water reservoir, as further described herein. In one embodiment, the external longitudinal fins 220 have upper horizontal ends 220a which terminate at or proximate to the underside or bottom of the larger diameter top portion 216 of the containment vessel 200, and lower horizontal ends 220b which terminate at or proximate to the base mat 304 of the concrete foundation 301. In one embodiment, the external longitudinal fins 220 may have a height H3 which is equal to or greater than one half of a total height of the shell of the containment vessel. In one embodiment, the upper horizontal ends 220a of the longitudinal fins 220 are free ends not permanently attached (e.g. welded) to the containment vessel 200 or other structure. At least part of the lower horizontal ends 220b of the longitudinal fins 220 may abuttingly contact and rest on a horizontal circumferential rib 222 welded to the exterior surface of the containment vessel shell 204 to help support the weight of the longitudinal fins 220 and minimize stresses on the longitudinal rib-to-shell welds. Circumferential rib 222 is annular in shape and may extend a full 360 degrees completely around the circumferential of the containment vessel shell 204. In one embodiment, the circumferential rib 222 is located to rest on the base mat 304 of the concrete foundation 301 which transfers the loads of the longitudinal fins 220 to the foundation. The longitudinal fins 220 may have a lateral extent or width that projects outwards beyond the outer peripheral edge of the circumferential rib 222. Accordingly, in this embodiment, only the inner portions of the lower horizontal end 220b of each rib 220 contacts the circumferential rib 222. In other possible embodiments, the circumferential rib 222 may extend (substantially) radially outwards far enough so that substantially the entire lower horizontal end 220b of each longitudinal rib 220 rests on the circumferential rib 222. The lower horizontal ends 220b may be welded to the circumferential rib 222 in some embodiments to further strengthen and stiffen the longitudinal fins 220. The external longitudinal fins 220 may be made of steel (e.g. low carbon steel), or other suitable metallic materials including alloys which are each welded on one of the longitudinally-extending sides to the exterior of the containment vessel shell 204. The opposing longitudinally-extending side of each rib 220 lies proximate to, but is preferably not permanently affixed to the interior of the inner shell 310 of the containment enclosure structure (CES) 300 to maximize the heat transfer surface of the ribs acting as heat dissipation fins. In one embodiment, the external longitudinal fins 220 extend (substantially) radially outwards beyond the larger diameter top portion 216 of the containment vessel 200 as shown. In one representative example, without limitation, steel ribs 220 may have a thickness of about 1 inch. Other suitable thickness of ribs may be used as appropriate. Accordingly, in some embodiments, the ribs 220 have a radial width that is more than 10 times the thickness of the ribs. In one embodiment, the longitudinal fins 220 are oriented at an oblique angle A1 to containment vessel shell 204 as best shown in FIGS. 2-3 and 5. This orientation forms a crumple zone extending 360 degrees around the circumference of the containment vessel 200 to better resist projectile impacts functioning in cooperation with the outer containment enclosure structure (CES) 300. Accordingly, an impact causing inward deformation of the containment enclosure structure (CES) shells 210, 211 will bend the longitudinal fins 220 which in the process will distribute the impact forces preferably without direct transfer to and rupturing of the inner containment vessel shell 204 as might possibly occur with ribs oriented 90 degrees to the containment vessel shell 204. In other possible embodiments, depending on the construction of the containment enclosure structure (CES) 300 and other factors, a perpendicular arrangement of ribs 220 to the containment vessel shell 204 may be appropriate. In one embodiment, referring to FIGS. 6-8, portions of the containment vessel shell 204 having and protected by the external (substantially) radial fins 220 against projectile impacts may extend below grade to provide protection against projectile strikes at or slightly below grade on the containment enclosure structure (CES) 300. Accordingly, the base mat 304 formed at the top of the vertically extending sidewalls 303 of the foundation 301 where the fins 220 terminate at their lower ends may be positioned a number of feet below grade to improve impact resistance of the nuclear reactor containment system. In one embodiment, the containment vessel 200 may optionally include a plurality of circumferentially spaced apart internal (substantially) radial fins 221 attached to the interior surface of the shell 204 (shown as dashed in FIGS. 2 and 3). Internal fins 221 extend (substantially) radially inwards from containment vessel shell 204 and longitudinally in a vertical direction of a suitable height. In one embodiment, the internal (substantially) radial fins 221 may have a height substantially coextensive with the height of the water-filled annulus 313 and extend from the base mat 304 to approximately the top of the shell 204. In one embodiment, without limitation, the internal fins 221 may be oriented substantially perpendicular (i.e. 90 degrees) to the containment vessel shell 204. Other suitable angles and oblique orientations may be used. The internal fins function to both increase the available heat transfer surface area and structurally reinforce the containment vessel shell against external impact (e.g. projectiles) or internal pressure increase within the containment vessel 200) in the event of a containment pressurization event (e.g. LOCA or reactor scram). In one embodiment, without limitation, the internal fins 221 may be made of steel. Referring to FIGS. 1-15, a plurality of vertical structural support columns 331 may be attached to the exterior surface of the containment vessel shell 204 to help support the diametrically larger top portion 216 of containment vessel 200 which has peripheral sides that are cantilevered (substantially) radially outwards beyond the shell 204. The support columns 331 are spaced circumferentially apart around the perimeter of containment vessel shell 204. In one embodiment, the support columns 331 may be formed of steel hollow structural members, for example without limitation C-shaped members in cross-section (i.e. structural channels), which are welded to the exterior surface of containment vessel shell 204. The two parallel legs of the channels may be vertically welded to the containment vessel shell 204 along the height of each support column 331 using either continuous or intermittent welds such as stitch welds. The support columns 331 extend vertically downwards from and may be welded at their top ends to the bottom/underside of the larger diameter top portion 216 of containment vessel housing the polar crane. The bottom ends of the support columns 331 rest on or are welded to the circumferential rib 222 which engages the base mat 304 of the concrete foundation 301 near the buried portion of the containment. The columns 331 help transfer part of the dead load or weight from the crane and the top portion 216 of the containment vessel 300 down to the foundation. In one embodiment, the hollow space inside the support columns may be filled with concrete (with or without rebar) to help stiffen and further support the dead load or weight. In other possible embodiments, other structural steel shapes including filled or unfilled box beams, I-beams, tubular, angles, etc. may be used. The longitudinal fins 220 may extend farther outwards in a (substantially) radial direction than the support columns 331 which serve a structural role rather than a heat transfer role as the ribs 220. In certain embodiments, the ribs 220 have a (substantially) radial width that is at least twice the (substantially) radial width of support columns. FIGS. 11-15 show various cross sections (both longitudinal and transverse) of containment vessel 200 with equipment shown therein. In one embodiment, the containment vessel 200 may be part of a small modular reactor (SMR) system such as SMR-160 by Htoltec International. The equipment may generally include a nuclear reactor vessel 500) with a reactor core and circulating primary coolant disposed in a wet well 504, and a steam generator 502 fluidly coupled to the reactor and circulating a secondary coolant which may form part of a Rankine power generation cycle. Other appurtenances and equipment may be provided to create a complete steam generation system. Referring primarily now to FIGS. 2-3, 16, and 18, the containment vessel 200 may further include an auxiliary heat dissipation system 340 including a plurality of internal longitudinal ducts 341 circumferentially spaced around the circumference of containment vessel shell 204. Ducts 341 extend vertically parallel to the vertical axis VA and in one embodiment are attached to the interior surface of shell 204. The ducts 341 may be made of metal such as steel and are welded to interior of the shell 204. In one possible configuration, without limitation, the ducts 341 may be comprised of vertically oriented C-shaped structural channels (in cross section) positioned so that the parallel legs of the channels are each seam welded to the shell 204 for their entire height to define a sealed vertical flow conduit. Other suitably shaped and configured ducts may be provided so long the fluid conveyed in the ducts contacts at least a portion of the interior containment vessel shell 204 to transfer heat to the water-filled annulus 313. Any suitable number and arrangement of ducts 341 may be provided depending on the heat transfer surface area required for cooling the fluid flowing through the ducts. The ducts 341 may be uniformly or non-uniformly spaced on the interior of the containment vessel shell 204, and in some embodiments grouped clusters of ducts may be circumferentially distributed around the containment vessel. The ducts 341 may have any suitable cross-sectional dimensions depending on the flow rate of fluid carried by the ducts and heat transfer considerations. The open upper and lower ends 341a, 341b of the ducts 341 are each fluidly connected to a common upper inlet ring header 343 and lower outlet ring header 344. The annular shaped ring headers 343, 344 are vertically spaced apart and positioned at suitable elevations on the interior of the containment vessel 200 to maximize the transfer of heat between fluid flowing vertically inside ducts 341 and the shell 204 of the containment vessel in the active heat transfer zone defined by portions of the containment vessel having the external longitudinal fins 220 in the primary annulus 313. To take advantage of the primary water-filled annulus 313 for heat transfer, upper and lower ring headers 343, 344 may each respectively be located on the interior of the containment vessel shell 204 adjacent and near to the top and bottom of the annulus. In one embodiment, the ring headers 343, 344 may each be formed of half-sections of steel pipe as shown which are welded directly to the interior surface of containment vessel shell 204 in the manner shown. In other embodiments, the ring headers 343, 344 may be formed of complete sections of arcuately curved piping supported by and attached to the interior of the shell 204 by any suitable means. In one embodiment, the heat dissipation system 340 is fluidly connected to a source of steam that may be generated from a water mass inside the containment vessel 200 to reject radioactive material decay heat. The containment surface enclosed by the ducts 341 serves as the heat transfer surface to transmit the latent heat of the steam inside the ducts to the shell 204 of the containment vessel 200 for cooling via the external longitudinal fins 220 and water filled annulus 313. In operation, steam enters the inlet ring header 343 and is distributed to the open inlet ends of the ducts 341 penetrating the header. The steam enters the ducts 341 and flows downwards therein along the height of the containment vessel shell 204 interior and undergoes a phase change from steam to liquid. The condensed steam drains down by gravity in the ducts and is collected by the lower ring header 344 from which it is returned back to the source of steam also preferably by gravity in one embodiment. It should be noted that no pumps are involved or required in the foregoing process. According to another aspect of the present disclosure, a secondary or backup passive air cooling system 400 is provided to initiate air cooling by natural convection of the containment vessel 200 if, for some reason, the water inventory in the primary annulus 313 were to be depleted during a thermal reactor related event (e.g. LOCA or reactor scram). Referring to FIG. 8, the air cooling system 400 may be comprised of a plurality of vertical inlet air conduits 401 spaced circumferentially around the containment vessel 200 in the primary annulus 313. Each air conduit 401 includes an inlet 402 which penetrates the sidewalls 320 of the containment enclosure structure (CES) 300 and is open to the atmosphere outside to draw in ambient cooling air. Inlets 402 are preferably positioned near the upper end of the containment enclosure structure's sidewalls 320. The air conduits 401 extend vertically downwards inside the annulus 313 and terminate a short distance above the base mat 304 of the foundation (e.g. approximately 1 foot) to allow air to escape from the open bottom ends of the conduits. Using the air conduits 401, a natural convection cooling airflow pathway is established in cooperation with the annulus 313. In the event the cooling water inventory in the primary annulus 313 is depleted by evaporation during a thermal event, air cooling automatically initiates by natural convection as the air inside the annulus will continue to be heated by the containment vessel 200. The heated air rises in the primary annulus 313, passes through the secondary annulus 330, enters the head space 318, and exits the dome 316 of the containment enclosure structure (CES) 300 through the vent 317 (see directional flow arrows, FIG. 8). The rising heated air creates a reduction in air pressure towards the bottom of the primary annulus 313 sufficient to draw in outside ambient downwards through the air conduits 401 thereby creating a natural air circulation pattern which continues to cool the heated containment vessel 200. Advantageously, this passive air cooling system and circulation may continue for an indefinite period of time to cool the containment vessel 200). It should be noted that the primary annulus 313 acts as the ultimate heat sink for the heat generated inside the containment vessel 200. The water in this annular reservoir also acts to maintain the temperature of all crane vertical support columns 331 (described earlier) at essentially the same temperature thus ensuring the levelness of the crane rails (not shown) at all times which are mounted in the larger portion 216 of the containment vessel 200. Operation of the reactor containment system 100 as a heat exchanger will now be briefly described with initial reference to FIG. 19. This figure is a simplified diagrammatic representation of the reactor containment system 100 without all of the appurtenances and structures described herein for clarity in describing the active heat transfer and rejection processes performed by the system. In the event of a loss-of-coolant (LOCA) accident, the high energy fluid or liquid coolant (which may typically be water) spills into the containment environment formed by the containment vessel 200. The liquid flashes instantaneously into steam and the vapor mixes with the air inside the containment and migrates to the inside surface of the containment vessel 200 sidewalls or shell 204 (since the shell of the containment is cooler due the water in the annulus 313). The vapor then condenses on the vertical shell walls by losing its latent heat to the containment structure metal which in turn rejects the heat to the water in the annulus 313 through the longitudinal fins 220 and exposed portions of the shell 204 inside the annulus. The water in the annulus 313 heats up and eventually evaporates forming a vapor which rises in the annulus and leaves the containment enclosure structure (CES) 300 through the secondary annulus 330, head space 318, and finally the vent 317 to atmosphere. As the water reservoir in annulus 313 is located outside the containment vessel environment, in some embodiments the water inventory may be easily replenished using external means if available to compensate for the evaporative loss of water. However, if no replenishment water is provided or available, then the height of the water column in the annulus 313 will begin to drop. As the water level in the annulus 313 drops, the containment vessel 200 also starts to heat the air in the annulus above the water level, thereby rejecting a portion of the heat to the air which rises and is vented from the containment enclosure structure (CES) 300 through vent 317 with the water vapor. When the water level drops sufficiently such that the open bottom ends of the air conduits 401 (see, e.g. FIG. 8) become exposed above the water line, fresh outside ambient air will then be pulled in from the air conduits 401 as described above to initiate a natural convection air circulation pattern that continues cooling the containment vessel 200. In one embodiment, provisions (e.g. water inlet line) are provided through the containment enclosure structure (CES) 300 for water replenishment in the annulus 313 although this is not required to insure adequate heat dissipation. The mass of water inventory in this annular reservoir is sized such that the decay heat produced in the containment vessel 200 has declined sufficiently such that the containment is capable of rejecting all its heat through air cooling alone once the water inventory is depleted. The containment vessel 200 preferably has sufficient heat rejection capability to limit the pressure and temperature of the vapor mix inside the containment vessel (within its design limits) by rejecting the thermal energy rapidly. In the event of a station blackout, the reactor core is forced into a “scram” and the passive core cooling systems will reject the decay heat of the core in the form of steam directed upper inlet ring header 343 of heat dissipation system 340 already described herein (see, e.g. FIGS. 16 and 18). The steam then flowing downwards through the network of internal longitudinal ducts 341 comes in contact with the containment vessel shell 204 interior surface enclosed within the heat dissipation ducts and condenses by rejecting its latent heat to the containment structure metal, which in turn rejects the heat to the water in the annulus via heat transfer assistance provide by the longitudinal fins 220. The water in the annular reservoir (primary annulus 313) heats up eventually evaporating. The containment vessel 200 rejects the heat to the annulus by sensible heating and then by a combination of evaporation and air cooling, and then further eventually by natural convection air cooling only as described herein. As mentioned above, the reactor containment system 100 is designed and configured so that air cooling alone is sufficient to reject the decay heat once the effective water inventory in annulus 313 is entirely depleted. In both these foregoing scenarios, the heat rejection can continue indefinitely until alternate means are available to bring the plant back online. Not only does the system operate indefinitely, but the operation is entirely passive without the use of any pumps or operator intervention. While the foregoing description and drawings represent some example systems, it will be understood that various additions, modifications and substitutions may be made therein without departing from the spirit and scope and range of equivalents of the accompanying claims. In particular, it will be clear to those skilled in the art that the present invention may be embodied in other forms, structures, arrangements, proportions, sizes, and with other elements, materials, and components, without departing from the spirit or essential characteristics thereof. In addition, numerous variations in the methods/processes described herein may be made. One skilled in the art will further appreciate that the invention may be used with many modifications of structure, arrangement, proportions, sizes, materials, and components and otherwise, used in the practice of the invention, which are particularly adapted to specific environments and operative requirements without departing from the principles of the present invention. The presently disclosed embodiments are therefore to be considered in all respects as illustrative and not restrictive, the scope of the invention being defined by the appended claims and equivalents thereof, and not limited to the foregoing description or embodiments. Rather, the appended claims should be construed broadly, to include other variants and embodiments of the invention, which may be made by those skilled in the art without departing from the scope and range of equivalents of the invention.
047088222	abstract	A method of solidifying radioactive waste wherein radioactive solid waste of a predetermined shape are embedded in a solidifying material which has a modulus of elasticity that is equal to, or smaller than, the modulus of elasticity of said waste, to provide a solidified body.. The modulus of elasticity of the solidifying material is adjusted to be equal to, or smaller than, that of the radioactive solid waste, in order to prevent stress concentrations at the boundaries between the solidifying material and the radioactive solid waste, particularly on the solidifying material side thereof. The invention makes it possible to prepare a solidified body with a desired durability and safety factor.
039986926	description	DESCRIPTION OF THE INVENTION The primary purpose of the pre-breeder reactor of the present invention is to produce fissile U.sup.233 for use in a breeder reactor. A reactor which accomplishes this is characterized by the following design parameters: 1. The U.sup.235 -enriched regions of the core should be designed such that the initial hydrogen-to-U.sup.235 atomic ratio is between 10.0 and 150.0 and the initial U.sup.235 to U.sup.238 ratio is between 0.005 and 0.20. 2. The plutonium-enriched regions of the core should be designed such that the initial plutonium-to-hydrogen atomic ratio is between 0.0 and 0.015 and the equilibrium ratio is between 0.002 and 0.015. The initial plutonium-to-Th.sup.232 ratio should be between 0.0 and 0.03 and the equilibrium ratio between 0.005 and 0.03. 3. The binary fuel compositions shall occupy from 30 to 95% of the core volume. 4. There will initially be at least two distinct fuel compositions. One composition shall contain thorium, possibly mixed with plutonium, and at least one composition shall contain a mixture of uranium with U.sup.235 as the principal fissile isotope and U.sup.238 as the diluent. 5. The fuel materials may be composed of (a) uranium or plutonium metal or metal alloys, (b) uranium or plutonium oxide with or without diluents of uranium oxide, thorium oxide, beryllium oxide or other low cross section oxides, (c) uranium or plutonium carbide with or without a diluent of uranium carbide, thorium carbide or a low cross section carbide diluent. 6. Whenever plutonium fuel is being used the isotopic mix may vary as the sources vary, but each plutonium composition must contain at least 60% of the fissile isotopes Pu.sup.239 or Pu.sup.241. With the use of both U.sup.235 --U.sup.238 and Pu.sup.239 --Th.sup.232 fuels, the reactor described herein may be employed under a wide variety of economic or technical constraints. If plutonium is not available in large quantities, then it would be possible to operate for a cycle with no plutonium in the thorium. In a later cycle the plutonium produced from the U.sup.235 of the first cycles could be used in the thorium fuel to improve conversion performance. Conversely, if a large amount of plutonium were available and the criteria of prime importance were the efficient utilization of plutonium already available rather than efficient over-all fuel utilization, then more plutonium could be used from the beginning of the first cycle of the pre-breeder to accelerate its conversion into a breeder. Thus, the invention describes a pre-breeder concept which is viable in almost any future economic environment. As shown herein, a combination of U.sup.235 and Pu.sup.239 as fissile materials and Th.sup.232 and U.sup.238 as fertile materials can be employed in such a way that 1. power comparable to a breeder is achieved PA1 2. lifetime comparable to a breeder is achieved PA1 3. conversion of Th.sup.232 to U.sup.232 is achieved PA1 4. some amount of makeup plutonium is achieved PA1 5. the materials are produced in core regions such that only chemical separation of materials is required PA1 6. The design of the pre-breeder is mechanically the same as that of the light-water breeder reactor mentioned hereinbefore and PA1 7. the use of different combinations of fuels can result in different fuel utilizations over a range of economic conditions. The advantages of this design are: 1. The invention provides an efficient method of generating U.sup.233 for use in light-water breeder reactors. 2. Fissile plutonium is used in the thorium fuel to enhance U.sup.233 conversion but the amount of plutonium used may be varied to strike any balance desired between over-all fuel utilization and the time required to achieve a full breeder reactor. 3. The invention provides a useful market for the plutonium currently being produced by light-water reactor plants without being dependent on that external plutonium source. 4. Because of its flexible plutonium usage, the invention can be economically optimized over a wide range of economic environments. 5. The invention has the capability of being used as a replacement core in existent light-water reactor plants. 6. Mechanically, the invention is of essentially the same design as light-water breeder reactors, simplifying conversion to breeding. To demonstrate the ability of the uranium-plutonium pre-breeder concept to achieve its goal of a high conversion ratio while producing a significant amount of power and attaining a reasonable lifetime, a typical design will be described. The typical design discussed herein is included only for illustrative purposes. It is not the only possible pre-breeder design, nor is it an optimal design since the possibility of a reduction in the fissile fuel loading or an improved conversion ratio exists. As shown in the drawing, a nuclear reactor incorporating the present invention comprises a core 10 enclosed within a pressure vessel 11 provided with a closure head 12. Core 10 is disposed within core cage assembly 13 which includes upper core barrel 14, lower core barrel 15 and bottom plate assembly 16. Inlet baffle shield barrel 17 is disposed between lower core barrel 15 and pressure vessel 11 and terminates in inlet flow baffle 18 having a plurality of openings 19 therein which allow coolant to circulate therethrough. Also forming a part of core cage assembly 13 is holddown barrel 20 which is inside of upper core barrel 14 and prevents upward movement of the reflector due to water flowing upwardly therethrough. The upper section of holddown barrel 20 and upper core barrel 14 contain slots 21 to provide for coolant exit from the core and to provide for differential thermal expansion. Pressure vessel 11 is provided with four inlet nozzles 22 communicating with a lower plenum 23 below and outside of baffle 18 and four outlet nozzles 24 communicating with an upper plenum 25 which communicates with slots 21. As more particularly shown in FIG. 2, core 10 includes a plurality of fuel modules 26, each including a hexagonal seed assembly 27 surrounded by a hexagonal annular blanket assembly 28. Reflector modules 29 surround blanket assemblies 28 to complete the core 10. Seed assemblies 27 are moved longitudinally with respect to blanket assemblies 28 by control mechanisms 29a (FIG. 1) to control the reactor. Blanket assemblies 28 are supported by support tubes 30 which are suspended from closure head 12 and seed assemblies 27 are supported by the control mechanism 29a through lead screws (not shown). Further details of the mechanical structure of this reactor will not be described because they form no part of the present invention. For these details reference is made to U.S. Pat. No. 3,957,575, supra. Each seed assembly 27 includes a plurality of parallel, elongated fuel rods 31 disposed in triangular array within an hexagonal shell 32 which separates coolant flow in the seed from coolant flow in the blanket. Each blanket assembly 28 includes a plurality of parallel, elongated fuel rods 31 disposed in triangular array in an hexagonal annulus around a seed assembly 27. Six rows of blanket fuel rods are required for the design specifically described herein. The fuel rods 31 in the blanket are supported by a blanket support tube 33 disposed interiorly of the blanket assembly which also serves to define a channel within which the seed assembly moves. These blanket fuel rods are restrained by a cage structure including corner posts 34 surrounding the blanket assembly as no shell is needed around the blanket assembly. As shown in FIG. 4, fuel rods 31 include an active portion 35 and axial blanket portions 36 and contain a plurality of fuel pellets 37 stacked inside of a hollow thin-walled cladding tube 38. The fuel rods in a seed assembly are 0.306 inch O.D. and the fuel rods in a blanket assembly are 0.571 inch O.D. The combined length of the active portion and axial blanket portions of a fuel rod 31 is 102 inches. The specific fuel loading employed in this embodiment of the invention employs oxide pellets of fuel making use of U.sup.235 as the fissile fuel in the seed fuel rods with U.sup.238 as fertile material and Pu.sup.239 as fissile fuel in the blanket fuel rods with Th.sup.232 as fertile material. The axial end blankets for the seed fuel rods contain natural uranium dioxide and the axial end blankets for the blanket fuel rods contain Th.sup.232 O.sub.2. Computation of the composition of fuel needed in the seed and blanket fuel rods was carried out assuming that the fissile-bearing fuel was 97% theoretical density and that the isotopic mix of the plutonium fuel was Pu.sup.239 78%, Pu.sup.240 12%, Pu.sup.241 10% plus a negligible trace of Pu.sup.242. The composition of the active portion of the fuel and blanket rods will next be given. The upper third of the active portion of the seed fuel rods is 1.458 weight percent U.sup.235, the middle third is 2.914 weight percent U.sup.235 and the lower third is 10.194 weight percent U.sup.235. Weight percent U.sup.235 is defined as ##EQU1## The active portion of the blanket fuel rods contains plutonium distributed as follows: The innermost row of blanket fuel rods 31 contains 0.473 weight percent Pu.sup.239, the second row contains 6.947 weight percent Pu.sup.239 in the upper five-sixths thereof and 0.473 weight percent Pu.sup.239 in the lower one-sixth thereof and the remaining four rows contain 2.362 weight percent Pu.sup.239 in the upper one-third thereof, 1.655 weight percent Pu.sup.239 in the next 1/6 thereof, 0.947 weight percent Pu.sup.239 in the next 1/3 thereof and 0.473 weight percent Pu.sup.239 in the bottom 1/6 thereof. Weight percent Pu.sup.239 is defined as ##EQU2## The fissile fuel loading of the reactor is 22 kilograms of U.sup.235 in the seed region and 10 kilograms of Pu.sup.239 in the blanket region of each module. The critical positions of the movable seed of the pre-breeder design are shown on FIG. 5 as a function of core lifetime. For comparative purposes, FIG. 5 also shows the critical positions for a typical light-water breeder reactor. Since the two designs are at the same power level, FIG. 5 indicates that the pre-breeder design is capable of essentially the same power lifetime at an LWBR design. In discussing the effectiveness of the typical pre-breeder design some figure of merit concerning the production of U.sup.233 must be defined. The main parameter of interest is the amount of U.sup.233 produced per net fissile atom lost over the entire core, which is defined at any time in life as: ##EQU3## where gain and loss at any time are measured from time O and (B) indicates the isotope is in the plutonium-thorium region and (S) indicates the U.sup.235 --U.sup.238 region. The isotope Pa.sup.233 is counted as U.sup.233 because it will decay to U.sup.233 with a 27.4-day half-life when removed from the core. The table below is a partial mass balance for the typical design at 4 times in life which also shows the appropriate C values. The value of C at 15,000 EFPH is 0.641. For this typical design, with assumed 1% fabrication and reprocessing losses, 5.21 pre-breeder cycles of 15,000 EFPH each are required to produce enough U.sup.233 for a light-water breeder reactor requiring 31.65 kg per module. Since in practice a full 6 cycles would probably be run, it would require 10.3 years to achieve a light-water breeder, but an additional 13% of 4.11 kg of U.sup.233 would have been produced. __________________________________________________________________________ TYPICAL PRE-BREEDER FISSILE FUEL MASS BALANCE __________________________________________________________________________ Effective Full Power Hours 0 3000 9000 15,000 __________________________________________________________________________ U.sup.233 .0108 .0095 .0079 .0067 SEED U.sup.235 22.041 20.943 18.569 16.381 (Movable Pu.sup.239 0 .6044 1.3943 1.9167 Region) Pu.sup.242 0 .0115 .0882 .2014 Pa.sup.233 0 .4886 .4954 .4816 BLANKET U.sup.233 0 1.0984 3.6276 5.5863 (Stationary U.sup.235 0 .0011 .0118 .0358 Region) Pu.sup.239 9.8390 7.9910 5.4596 3.7094 Pu.sup.241 1.2720 1.3175 1.4172 1.4380 C -- 0.695 0.664 0.641 __________________________________________________________________________
052326578	claims	1. In a nuclear fuel storage body including a pool of fast neutron slowing fluid and a plurality of upright storage cans submerged in said fluid and disposed in a spaced side-by-side array, each storage can being composed of a plurality of side walls connected together to receive and store a nuclear fuel assembly, a plurality of flux trap neutron absorber arrangements disposed in said storage body between said storage cans, each of said flux trap neutron absorber arrangements comprising: (a) separate means extending vertically along and attached to the exterior of each of adjacent side walls of adjacent pairs of said spaced storage cans for forming respective pockets extending along said adjacent side walls and being spaced from one another, said pocket forming means being an outer elongated covering extending vertically along and attached along its periphery to each of said adjacent side walls; (b) an elongated flat plate of a thermal neutron absorber material mounted in each of said pockets, said plates of thermal neutron absorber material being likewise spaced from one another and defining a fast neutron slow-down region therebetween; (c) a slab of a metal hydride disposed in said fast neutron slow-down region between said plates of thermal neutron absorber material and said separate pocket forming means on said adjacent side walls; and (d) a canister containing said slab of said metal hydride being disposed in said fast neutron slow-down region, said canister being connected to one of said adjacent side walls of said adjacent storage cans; (e) said canister including a hollow metal container having a rectangular cross-section defined by a pair of spaced side walls and a pair of lateral end walls extending between and rigidly interconnecting said side walls, said canister being filled by said slab of metal hydride which is disposed in contact with said side and lateral end walls of said canister, said side walls of said canister being disposed in contact with said adjacent storage cans such that no gap and thereby no fast neutron flowing fluid is present between said canister and said adjacent storage cans, (f) said canister further including a pair of lateral mounting flanges attached to and extending in opposite directions outwardly from opposite vertical edges of one of said side walls, said lateral mounting flanges being attached to opposite vertical edge portions of said respective covering. 2. The arrangement as recited in claim 1, wherein said outer covering is a plate of metal. 3. The arrangement as recited in claim 1, wherein each of said storage cans has four of said side walls, four of said canisters being respectively disposed adjacent said four side walls of said storage can, two of said four canisters being mounted by the two adjacent side walls of said four side walls of said storage can and the other two of said four canisters disposed in abutting contact with the two other adjacent side walls of said storage can. 4. The arrangement as recited in claim 1, wherein said thermal neutron absorber material of said plates is boron carbide. 5. The arrangement as recited in claim 1, wherein said metal hydride is titanium hydride.
043615341	summary	This invention concerns the simultaneous measurement of the chemical concentrations of the silicon and aluminium constituents of materials. It provides both a method and apparatus for that purpose. There are many possible applications of the present invention, including the measurement of silicon and aluminium in coal and in iron ores. However, the invention was developed primarily to permit the monitoring of the chemical concentrations of aluminium and silicon in bauxite ores, as part of a quality control process for the mineral industry, and it is this application of the invention that will be described in detail in this specification. There are two particular areas in the bauxite industry where the present invention will be used. One is the monitoring of ore quality during ship-loading operations of bauxite for export, where monitoring of aluminium grade and silicon impurity concentrations are essential to ensure that the ore satisfies export contract specifications. The other use is the monitoring of ore quality whilst sorting the bauxite into stockpiles of different specified chemical concentrations of silicon and aluminium. Depending on the way in which they have been formed, these stockpiles (a) may contain ore which has been blended for ore treatment plants or (b) may be used in subsequent blending operations. In both of these situations, the current practice in monitoring the ore quality involves the periodic sampling of the ore from the bulk supply, which is usually moving on a conveyor belt when the sample is taken. The samples are moderately large (several kilogrammes) and are either subsampled immediately, or are mixed with other samples, taken by a predetermined number of automatic sampling cycles to form a representative bulk sample, which is then sub-sampled. Sub-sampling and crushing proceeds until a small specimen (of about 1 g) is ultimately available for chemical analysis by wet chemical assaying or by X-ray fluorescence analysis procedures. These sample preparation procedures are particularly time-consuming if good representivity of the bulk is required in the sample. The analysis is also time-consuming. It has been found that in some situations (for instance during ship loading), when variations of ore quality occur these analytical methods are not fast enough to permit steps to be taken to correct the chemical concentrations of aluminium and silicon (for example, by further blending measures). If it were possible to apply the prior art techniques to on-stream analysis of bauxite on a moving belt, a more rapid analysis of the constituents and hence more rapid corrective blending measures should, in principle, be possible. Unfortunately, wet chemical methods cannot be applied to an on-stream situation, and on-stream X-ray fluorescence methods are inapplicable to lump-flow measurement. The X-ray fluorescence method is also unsuitable for the analysis of untreated bulk samples, due to the low penetration of X-radiation (less than 1 mm), and the fact that the ore is heterogeneous as regards moisture and particle size. Neutron activation analysis, which is the basis of the present invention, does not have the problems associated with wet chemical assaying or the X-ray fluorescence technique, noted above, when applied to the analysis of large bulk samples. Indeed, activation analysis methods are directly applicable to large bulk samples and require minimal sample preparation in terms of crushing and drying. They also avoid most of the heterogeneity problems associated with the application of X-ray fluorescence analysis to bulk samples because the neutrons and gamma rays involved have a much deeper penetration than X-rays. For this reason, the monitoring of bauxite ore quality (and the aluminium and silicon content of other materials) on moving belts is amenable to the neutron activation method of the present invention. Neutron activation methods have previously been applied to the analysis of silicon and aluminium in small samples (less than 150 g). For example, they have been described in the paper by F. Dugain and J. Tatar in Ann. Inst. Geol. Publici Hungary, Volume 54, p 375 (1970), and in the papers by L. Alaerts, J. P. Op de Beeck, and J. Hoste, in Anal. Chim. Acta, Volume 70, p 253 (1974) and in Anal. Chim. Acta, Volume 78, p 329 (1975). These methods depend on two interactions with the constituent chemical elements. One interaction is that occurring between fast neutrons and .sup.28 Si, producing .sup.28 Al by the reaction .sup.28 Si(n,p).sup.28 Al. The product, .sup.28 Al, decays with a 2.3 minute half life, emitting 1.78 MeV gamma radiation. Similarly, when the aluminium constituent of the sample is irradiated with slow neutrons, the same radioactive isotope, .sup.28 Al, is produced with the consequent emission of 1.78 MeV gamma radiation. Since there is negligible interaction between the fast neutrons and the material in small samples, the chemical concentrations of silicon and aluminium can be calibrated directly against the number of 1.78 MeV gamma ray counts observed within a given time interval after irradiation first with fast neutrons and then with slow neutrons. With bulk samples of bauxite, particularly samples having a significant water content, special allowance must be made for the slowing down of the fast neutrons used for the silicon analysis and the associated production of .sup.28 Al due to the capture of slow neutrons by the aluminium during the same irradiation. One method of allowing for this effect is described below, in the description of the operation of the present invention. For the currently used neutron activation analysis techniques, because two irradiations are necessary (with fast, then slow neutrons), there is a considerable capital outlay on the two neutron sources and their respective shielding assemblies. If the analysis is applied to a moving stream of ore on a belt, two spectrometric gamma ray detectors [for example, 127.times.127 mm NaI(T1)] are also required. If the analysis is performed on bulk samples contained in bins or boxes, although only one gamma ray detector is necessary, the analysis procedures for silicon and aluminium must duplicate each other, which doubles the necessary time and effort for analysis. In addition, for the measurements to be useful to the analyst, it is essential that the fast and slow neutron flux should be constant and reproducible from one measurement to the next. The present invention offers appreciable savings in time and equipment cost, compared with current technology, by providing a method of analysis which is based on a single sample irradiation followed by a single measurement procedure. The nuclear reactions providing the basis of the present invention are: .sup.27 Al(n,p).sup.27 Mg (used for the determination of the aluminium constituent), and .sup.28 Si(n,p).sup.28 Al (for the silicon determination). The energies of neutrons effective in these reactions are greater than 4.5 MeV. The radioactive nucleus .sup.27 Mg decays with a half life of 9.46 minutes and emits two gamma rays during its decay, which have energies of 0.844 MeV and 1.055 MeV respectively. The emission and half life of the other radioactive nucleus, .sup.28 Al, have been described above. As previously mentioned, a third nuclear reaction is important with all bulk samples having significant water content as well as aluminium and silicon constituents. This reaction, .sup.27 Al(n,.gamma.).sup.28 Al, which entails the capture of slow neutrons in aluminium, results in the emission of 1.78 MeV gamma radiation which is additional to the 1.78 MeV gamma radiation resulting from the fast neutron reaction with the silicon constituent of the sample. (Note that even with sources emitting only fast neutrons for sample irradiations, appreciable numbers of fast neutrons are slowed down to thermal energies within the sample by their collision with the hydrogen nuclei associated with the water content of the sample). Applications of the above nuclear reactions for the fast neutron activation analysis of aluminium and silicon have been described in the scientific literature. For example, reference can be made to the paper by R. H. Gijbels and J. Hertogen in Pure Appl. Chem., Volume 49, p 1555, (1977), and the paper by J. Kuusi in Nucl. Appl. Technol., Volume 8, p 465 (1970). However, these applications are either for small samples, or for larger samples that contain little hydrogen and therefore cause negligible moderation of the fast neutrons within the samples. The present invention overcomes the problem of interference by aluminium with silicon from the 1.78 MeV gamma radiation in the following way. After the fast neutron irradiation, the 1.78 MeV gamma rays from the sample are measured concurrently with those emitted by .sup.27 Mg at 0.844 MeV and 1.015 MeV for a preset time interval. (If the sample container is fabricated from a material such as copper, which produces gamma radiation interfering with the 1.015 MeV gamma rays of the sample, the 1.015 MeV gamma rays are excluded from the analysis). Since the number of counts from .sup.27 Mg are due only to aluminium, the chemical concentration of aluminium can be related directly to these recorded counts, given a knowledge of the mass of the sample. With a knowledge of the aluminium content of the sample, provided the slow neutron flux in the material is also known, the component of the 1.78 MeV gamma radiation due to slow neutron reactions with the aluminium can be subtracted from the total 1.78 MeV gamma radiation count to provide the gamma radiation at 1.78 MeV resulting from fast neutron activation of the silicon in the sample. Because the thermal neutron flux within the bulk sample is sensitive to water content, it is essential to measure the number of thermal neutrons in a given time interval during the neutron irradiation. For this purpose, a suitable neutron detector will be located adjacent to the sample. The number of neutrons recorded by the detector is proportional to the thermal neutron flux within the sample. Thus, according to the present invention, a method of simultaneously analysing the aluminium and silicon content of a sample of material comprises the steps of: (a) irradiating the sample with fast neutrons; (b) monitoring the thermal neutron flux within the sample; (c) monitoring the gamma radiation from the irradiated sample at energies of 1.78 MeV and 1.015 and/or 0.844 MeV; (d) using the monitored gamma radiation at 1.015 and/or 0.844 MeV to estimate the aluminium content of the sample; and (e) using the monitored gamma radiation at 1.78 MeV, compensated by the gamma radiation at 1.78 MeV due to the thermal neutron reaction with the estimated aluminium in the sample, to estimate the silicon content of the sample. Also according to the present invention, apparatus for the simultaneous analysis of aluminium and silicon content of a sample of material comprises: (a) a fast neutron source, adapted to irradiate the sample of material; (b) a thermal neutron detector, located to monitor the thermal neutron flux in the irradiated sample; and (c) a gamma ray detector separated from the neutron source and shielded therefrom, adapted to monitor the gamma spectrum from the irradiated sample, at least at energies of 1.78 MeV and of 0.844 and/or 1.015 MeV.
	abstract	Certain exemplary embodiments of the present invention comprise a device comprising a cast computed-tomography collimator descended from a lithographically-derived micro-machined metallic foil stack lamination mold. Certain exemplary embodiments of the present invention comprise a method comprising filling a mold having a stacked plurality of micro-machined metallic foil layers with a first casting material to form a first cast product; demolding the first cast product from the mold; filling the first cast product with a second casting material to form a cast computed-tomography collimator; and demolding the cast computed-tomography collimator from the first cast product. It is emphasized that this abstract is provided to comply with the rules requiring an abstract that will allow a searcher or other reader to quickly ascertain the subject matter of the technical disclosure. This abstract is submitted with the understanding that it will not be used to interpret or limit the scope or meaning of the claims.
054854962	description	DETAILED DESCRIPTION The absorbable biomaterials that constitute all or part of the medical devices or products that are advantageously treated by the invention herein are biodegradable polymers and copolymers without gamma irradiation stabilizers incorporated into the chemical structure. These include (1) biodegradable linear aliphatic homopolymer and copolymer polyesters; (2) biodegradable copolymers formed by copolymerizing (a) monomers which polymerize to form linear aliphatic polyesters with (b) monomers which do not polymerize to form linear aliphatic polyesters, or formed to be block copolymers of (a) and (b); and (3) biodegradable polymers and copolymers which do not include gamma irradiation stabilizing moieties other than (1) and (2). The weight average molecular weights of these polymers and copolymers typically range from 10,000 to 500,000, preferably from 20,000 to 125,000. Examples of biodegradable linear aliphatic homopolymer polyesters include poly(alpha-hydroxy C.sub.1 -C.sub.5 alkyl carboxylic acids), e.g., polyglycolic acids (e.g., sold under the tradename Dexon), poly-L-lactides, and poly-D,L-lactides; poly-3-hydroxy butyrate; polyhydroxyvalerate; polycaprolactones, e.g., poly(epsilon-caprolactone); polyglutamates; and modified poly(alpha-hydroxyacid)homopolymers, e.g., homopolymers of the cyclic diester monomer, 3-(S)[alkyloxycarbonyl)methyl]-1,4-dioxane-2,5-dione which has the formula 4 where R is lower alkyl, depicted in Kimura, Y., "Biocompatible Polymers", in Biomedical Applications of Polymeric Materials, Tsuruta, T., et al, eds., CRC Press, 1993 at page 179. Examples of biodegradable linear aliphatic copolymer polyesters are glycolide-lactide copolymers (e.g., sold under the trademark Vicryl), glycolide-caprolactone copolymers, poly-3-hydroxy butyrate-valerate copolymers, and copolymers of said cyclic diester monomer, 3-(S)[(alkyloxycarbonyl)methyl]-1,4-dioxane-2,5-dione, with L-lactide. The glycolide-lactide copolymers include poly(glycolide-L-lactide) copolymers formed utilizing a monomer mole ratio of glycolic acid to L-lactic acid ranging from 5:95 to 95:5 and preferably a monomer mole ratio of glycolic acid to L-lactic acid ranging from 45:65 to 95:5, e.g., a monomer mole ratio of glycolic acid to L-lactic acid of 90:10 or a monomer mole ratio of glycolic acid to L-lactic acid of 50:50. The glycolide-caprolactone copolymers include glycolide and epsilon-caprolactone block copolymer, e.g., Monocryl or Poliglecaprone. Examples of biodegradable copolymers formed by copolymerizing (a) monomers which polymerize to form linear aliphatic polyesters with (b) monomers which do not polymerize to form linear aliphatic polyesters or formed to be block copolymers of (a) and (b) include poly(L-lactic acid-L-lysine) as described in Barrera, D. A., et al, JACS, Vol. 115, pp. 11010 to 11011 (1993); tyrosine based polyarylates, tyrosine-based polyiminocarbonates and tyrosine-based polycarbonates as described in Kohn, J., The 20th Annual Meeting of the Society of Biomaterials Apr. 5-9, 1994 publication, page 67 and in Nathan, A., et al, "Amino Acid Derived Polymers" in Biomedical Polymers, edited by Shalaby, S. W., pages 128, 129, Hauser Publishers, New York, 1994; poly(D,L-lactide-urethanes) as described in Storey, R. F., et al, ANTEC '92, 734-737 (1992); poly(ester-amides) as described in Barrows, T. H., "Bioabsorable Poly(ester-amides)", in Biomedical Polymers, edited by Shalaby, S. W., pages 100-101, Hauser Publishers, New York, 1994; and glycolide and trimethylene carbonate block copolymer, e.g., Maxon. Examples of other biodegradable polymers and copolymers that constitute materials of construction for medical devices or products advantageously treated by the invention herein include poly[bis(carboxylatophenoxy)phosphazene] as described in Cohen S., JACS, 112, 7832-7833 (1990); polyanhydrides, e.g., polymaleic anhydride, polysuccinic anhydride and polyglutaric anhydride; polycyanoacrylates, e.g., poly(alkyl-alpha-cyanoacrylate); and poly-p-dioxanone, e.g., PDS-II. The non-absorbable biomaterials that constitute materials of construction for medical devices or products advantageously treated herein include ultra high molecular polyethylene, e.g., polyethylene of weight average molecular weight ranging from 1.times.10.sup.6 to 7.times.10.sup.6, and polypropylene. We turn now to the provision of the substantial absence of oxygen as required in the invention herein. This is readily carried out by positioning the medical device or product to be sterilized in a gamma irradiation transparent container, e.g., a Pyrex container, and applying vacuum, e.g., using a diffusion pump, to lower the pressure to 1.times.10.sup.-5 to 1.times.10.sup.-7 torr, preferably to about 5.times.10.sup.-6 torr, and then sealing the container, e.g., using a flame. Less preferably, the substantial absence of oxygen can be provided by positioning the medical device or product to be sterilized in a gamma irradiation transparent container and replacing air in the container with an inert gas, e.g., nitrogen, argon or helium. We turn now to the requirement of a temperature ranging from -180.degree. C. to -200.degree. C. This is readily carried out by utilizing liquid nitrogen or liquid argon, e.g., by immersing oxygen depleted atmosphere container with the medical device or product therein in a gamma irradiation transparent Dewar flask or by spraying the liquified cooling gas on said container. We turn now to the gamma irradiation treatment. This is readily carried out by positioning the medical device or product to be sterilized in a gamma irradiation sterilizing apparatus relying on Co.sup.60 as a source of gamma irradiation. The medical device or product can be in an oxygen depleted atmosphere container in a Dewar of liquified cooling gas or the gamma irradiation sterilizing apparatus can be modified to so it contains a substantially oxygen free atmosphere (e.g., a vacuum chamber) containing a bath of liquified cooling gas or means to spray said liquified cooling gas on the device or product being sterilized. Gamma irradiation is typically applied at a rate of 0.38 to 0.45 Mrad/hr. and a sterilizing total dosage is often considered to be 2 Mrad and typically ranges from 2 Mrad to 4 Mrad. The invention is illustrated by the following Examples. In Examples I and II commercial medical products were utilized which had already been sterilized using ethylene oxide. The ethylene oxide treatment does not affect the strength properties of the medical products so the results of Examples I and II should be the same as if the gamma irradiation were the only sterilizing treatment. EXAMPLE I The manufacturer's sealed packs of 2-0 size absorbable suture thread composed of Dexon were opened and 5 inch lengths were cut and each 5 inch length was placed in a 3 mm/5 mm (i.d./o.d) Pyrex tube. Dry vacuum was drawn on each tube with a diffusion pump down to 5.times.10.sup.-6 torr. Then each tube was flame sealed. The Pyrex tubes with suture thread therein were immersed in a Dewar containing liquid nitrogen (77.degree. K. or -196.degree. C.). The suture material containing tubes immersed in the Dewar were gamma irradiated at a dosage rate of 0.45 Mrad per hour with a total dosage of 2 Mrads or 4 Mrads. After irradiation was carried out, the tubes were removed from the liquid nitrogen and were placed in ambient surroundings overnight whereby they came to room temperature. Each tube was then broken and the irradiated suture material was removed and placed in a phosphate buffer solution (pH 7.44) and 37.degree. C. for 3, 7, 10, 14, 17 or 21 days. The phosphate buffer solution at 37.degree. C. simulated the in vivo environment envisioned for the medical device. The buffer solution was replaced every 2 days to ensure constant pH. Suture material samples removed from the buffer solution were vacuum dried at 25.degree. C. overnight and then tensile breaking force was measured under ASTM Testing condition D1776-74 (21.degree. C. and 65% relative humidity) using an Instron Universal Testing Machine with a crosshead speed of 10 mm/min. The controls were processed the same but without the application of vacuum or immersion in liquid nitrogen, i.e., the suture material in the tubes was in contact with air at room temperature during gamma irradiation. FIG. 1 shows the results for the sutures gamma irradiated with a total dose of 2 Mrad, and FIG. 2 shows the results for the sutures gamma irradiated with a total dose of 4 Mrad. The filled in rectangles are the results for the samples maintained under vacuum and in liquid nitrogen during gamma irradiation. The plus signs are the results for the samples maintained in air at 30.degree. C. during gamma irradiation. EXAMPLE II The procedure of Example I was followed except that the absorbable suture thread was composed of Vicryl. FIG. 3 shows the results for the sutures gamma irradiated with a total dose of 2 Mrad, and FIG. 4 shows the results for the sutures gamma irradiated with a total dose of 4 Mrad. The filled in rectangles are the results for the samples maintained under vacuum and in liquid nitrogen during gamma irradiation. The plus signs are the results for the samples maintained in air at 30.degree. C. during gamma irradiation. EXAMPLE III When suture material composed of PDSII is substituted for the suture material of Dexon in Example I, significantly increased resistance to strength loss over time in an in vitro environment simulating an in vivo environment, is obtained for samples gamma irradiated under vacuum and liquid nitrogen compared to controls gamma irradiated in air at 30.degree. C. EXAMPLE IV When suture material composed of Maxon is substituted for the suture material of Dexon in Example I, significantly increased resistance to strength loss over time in an in vitro environment simulating an in vivo environment, is obtained for samples gamma irradiated under vacuum and liquid nitrogen compared to controls gamma irradiated in air at 30.degree. C. EXAMPLE V When suture material composed of Monocryl is substituted for the suture material of Dexon in Example I, significantly increased resistance to strength loss over time in an in vitro environment simulating an in vivo environment, is obtained for samples gamma irradiated under vacuum and liquid nitrogen compared to controls gamma irradiated in air at 30.degree. C. EXAMPLE VI A tibia component of a joint prostheses, composed of ultrahigh weight average molecular weight polyethylene is placed in a gamma irradiation transparent container which is evacuated to remove oxygen and then is immersed in liquid nitrogen. The tibia component in the evacuated container under liquid nitrogen is gamma irradiated at 0.45 Mrad/hr. for a total dose of 2 Mrad providing sterility. In another case, gamma irradiation is carried out in air at room temperature. The tibia component irradiated under vacuum and liquid nitrogen has significantly increased resistance to strength loss over time in an in vitro environment simulating an in vivo environment, compared to the tibia component gamma irradiated in air at 30.degree. C. EXAMPLE VII Ethylene oxide sterilized absorbable sutures composed of Dexon were immersed in a suspension of Staphylococcus epidermidis (1.times.10.sup.6 cells/ml) in phosphate buffer solution at 37.degree. C. for about 15 minutes. The sutures wee removed from immersion in the suspension and excess liquid was separated. The bacterial contaminated sutures were placed in Pyrex tubes (5 per tube) and dry vacuum was drawn on each tube overnight down to 1.times.10.sup.-5 torr. The tubes were sealed under vacuum. The sealed tubes were immersed in liquid nitrogen. The sealed tubes at liquid nitrogen temperature were gamma irradiated with a total dosage of 2.5 Mrad. After irradiation was carried out, the tubes were removed from liquid nitrogen and were placed in ambient surroundings whereby they came to equilibrium. Then the tubes were placed inside a septic hood and opened and the sutures were removed. Of the 5 sutures from each tube, two were cultured in a bacterial culture plate wherein the bacterial culture medium was trypticase soy agar with 5% sheep blood, two were cultured in trypticase soy broth and one was reserved for scanning electron microscope examination. The culture plates and the broth with sutures therein were incubated overnight whereupon examination was carried out for bacterial growth. Six tubes were prepared and processed in this way. Two controls were utilized. In one case, one tube was processed as above except that liquid nitrogen immersion and gamma irradiation were not carried out. In the second case, one tube was processed as above except that vacuum treatment, liquid nitrogen immersion and gamma irradiation were not carried ut. In all the cases where the invention was utilized, there was no bacterial growth. In the control case where vacuum was utilized, bacterial growth was noted on one of the two sutures incubated in the bacterial culture plate and on one of the two sutures incubated in trypticase soy broth. In the control case where vacuum was not utilized, bacterial growth was noted on both sutures incubated in the bacterial culture plate and on both sutures incubated in trypticase soy broth. Variations will be evident to those skilled in the art. Therefore the scope of the invention is determined by the claims.
	abstract	Various aspects of the present invention pertain to methods of sorption of various materials from an environment, including radioactive elements, chlorates, perchlorates, organohalogens, and combinations thereof. Such methods generally include associating graphene oxides with the environment. This in turn leads to the sorption of the materials to the graphene oxides. In some embodiments, the methods of the present invention also include a step of separating the graphene oxides from the environment after the sorption of the materials to the graphene oxides. More specific aspects of the present invention pertain to methods of sorption of radionuclides (such as actinides) from a solution by associating graphene oxides with the solution and optionally separating the graphene oxides from the solution after the sorption.
	description	The present invention relates to a multi-leaf collimator. Multi-leaf collimators (MLC) are used (principally) in the field of radiotherapy. A beam of radiation is directed toward a patient and must be collimated to fit the shape of the area to be treated. It is important to ensure that the dose in the areas outside that shape is as low as possible, but also that the whole area is treated. If areas are left untreated then the likelihood of recurrence is increased, whereas if non-treatment regions are irradiated then damage will be caused to healthy tissue resulting in greater side effects and longer recovery times after treatment. As the treatment area is rarely rectilinear, multi-leaf collimators are employed. These comprise an array of finger-shaped leaves of a radiation-absorbing material, each disposed in a parallel relationship and each able to move longitudinally relative to the others. By moving each leaf to a selected position, a collimator is provided which can exhibit a non-linear edge. In general, one such array (or “bank”) will be provided on each side of the beam. Multi-leaf collimators generally suffer from two difficulties. One is the leakage of radiation between the leaves of the bank, and the other is that the leaves often have square ends and thus exhibit a pixellated pattern when aligned to an edge that is non-perpendicular to the leaf direction. Various designs are employed to resolve the leakage rate, including stepped edges to the leaves, which therefore interlock (to an extent) and limit the clear view between leaves. However, inter-leaf leakage is a limiting factor in some treatment plans such as Intensity Modulated Radiation Therapy (IMRT) where the treatment time is relatively long. The pixellation issue is a factor related to the resolution of the leaf bank, and therefore efforts to alleviate this problem tend to involve the use of narrower leaves. This does, however, make the collimator more complex and present significant engineering challenges. JP03009767 suggests the use of two banks of leaves, offset so that the leaves of one cover the gaps between leaves of the other. This results in an improved resolution and reduced leakage, but still gives a pixellated effect. The present invention seeks to provide a multi-leaf collimator which further alleviates the problems of inter-leaf leakage and pixellation and provides improved resolution and large field size. According to a first aspect of the invention there is provided a collimator for a radiation beam comprising a first multi-leaf collimator set, a second multi-leaf collimator set at an acute angle to the first, and a third multi-leaf collimator set at an acute angle to the second. According to a second aspect of the invention there is provided a collimator for a radiation beam comprising a first multi-leaf collimator set of a certain depth, a second multi-leaf collimator set of a depth which is less than the depth of the first set and set at an acute angle to the first, and a third multi-leaf collimator set of a depth which is less that the depth of the second set and which is set at an acute angle to the second. Each multi-leaf collimator set will usually include a pair of leaf banks mutually opposed to each other. The acute angle between the first and the second multi-leaf collimator set is preferably the same as the acute angle between the second and the third set. A suitable angle is approximately 60°, however other suitable angles may also be used. The penumbra is the region close to the radiation field edge. The penumbral width is typically defined by the distance between the points where 20% and 80% of the dose at central axis is delivered. A small penumbra means a good beam definition which allows to give maximum dose to a target volume with a rapid dose fall off next to it towards the surrounding normal healthy tissue. A number of design features in such a collimator are preferred in order to improve the penumbra characteristics: The leaves of the multi-leaf collimator closest to the radiation source can be deeper in the direction of the radiation than the leaves of a multi-leaf collimator more distant from the radiation source. The leaves of the multi-leaf collimator furthest from the radiation source can be shallower in the direction of the radiation than the leaves of a multi-leaf collimator closer to the radiation source. The tips of the leaves of the multi-leaf collimators can be rounded. The radius of curvature of the tips of the leaves of the multi-leaf collimator closest to the radiation source can be greater than the radius of curvature of the tips of the leaves of a multi-leaf collimator more distant from the radiation source. The radius of curvature of the tips of the leaves of the multi-leaf collimator furthest from the radiation source can be less than the radius of curvature of the tips of the leaves of a multi-leaf collimator closer to the radiation source. In general, it is also preferred that the first multi-leaf collimator is closest to the radiation source, the third multi-leaf collimator is furthest from the radiation source, and the second multi-leaf collimator is between the first and third multi-leaf collimators. FIG. 1 shows the relevant part of a beam collimator 1 of conventional design. The collimator 1 comprises a set of a pair of multi-leaf banks 2, 2a disposed in a plane orthogonal to the direction of the X-ray or other beam 3 exiting from the aperture 4. FIG. 2 shows the collimator 10 according to the present invention, in a corresponding vertical plane (i.e. in the direction of the beam 3) to FIG. 1 of the conventional type of collimator. The collimator comprises three multi-leaf sets, 11a, 11b and 11c, each set comprising a multi-leaf bank disposed on either side of the X-ray beam 3, and each disposed in a row in the vertical section. What is not visible in FIG. 2 is that the three banks are arranged in different orientations, such as with their leaf direction at 60° to each other. FIG. 10 shows a top view of the multi-leaf collimator 10 of FIG. 2, wherein FIG. 10 illustrates an example of acute angles of 60 degrees between leaf directions of the sets 11a, 11b and between leaf directions of the sets 11b and 11c. FIG. 3 shows a typical clinical sample 20 with an irregular, rectilinear shape, which is to be treated using treatment plans such as IMRT. The area inside the solid black line forming the perimeter of the shape is the tissue which must be treated with X-rays or other radiation. In order that the tissue surrounding the treatment area, i.e. the healthy tissue, is not affected, the area to be treated must receive the radiation treatment whilst the healthy tissue must be shielded from the radiation using suitable radiation-blocking material. FIG. 4 shows the achieved irradiation pattern using a two-bank multi-leaf collimator such as the conventional type shown in JP03009767. The width of the leaves in this particular example are of 10 mm. The width of the leaves is measured in the plane orthogonal to the direction of the beam. As can be seen, large areas of healthy tissue will be irradiated due to the approximate boundary that is achieved using the conventional two-bank multi-leaf collimator. If the width of the leaves is decreased, i.e. to achieve a better resolution, in this example to 4 mm to produce a mini-collimator, the resulting irradiation pattern is closer to the pattern of the tissue to be treated, as shown in FIG. 5. Of course, the narrower the leaves of the bank, the better the resolution, however, the leaves of the bank can only be narrowed to a finite width as engineering challenges become insurmountable. Furthermore, the narrower the leaves of the bank, the smaller becomes the practically achievable field size, thus making it difficult to treat large areas. The field size of the conventional two bank multi-leaf collimator is of the order of 40×40 cm2. The field size of the mini-collimator is of the order of 16×20 cm2. The field size of the collimator according to the present invention, however, is at least 40 cm diameter, and may be greater. FIG. 6 shows the achieved irradiation pattern using a collimator according to the present invention, with 10 mm wide leaves. As can be seen the target irradiation pattern is improved substantially over that achieved by using either the 10 mm width leaves of the conventional two-bank multi-leaf collimator or the narrower leaves of the mini-collimator used to achieve the pattern in FIG. 5. FIG. 7 shows the overdose, i.e. irradiation of healthy tissue, applied for a standard target and range of orientations. As can be seen, using the conventional two-bank collimator, the overdose is minimum in only two orientations, i.e. when the area to be treated has a vertical or horizontal edge and thus the leaves of the conventional collimator may be placed in a close abutting relationship with the edge. The same is true of the mini-collimator, where again the overdose forms minima at two orientations. The results of the collimator according to the present invention, show a marked and substantial improvement in the reduction of the overdose, as can be seen from the graph, most of the overdose line lies well below the line achieved using the mini-collimator. FIG. 8 shows the embodiment according to a second aspect of the invention. The multi-leaf collimator set is shown partially fitted into an accelerator. The accelerator comprises a target to generate the radiation, a primary collimator 31 which partially collimates the beam 3, and filters 32 to filter the beam. The beam then passes through the multi-leaf collimator according to a second aspect of the invention. The depth of the leaves is measured in a direction parallel to the direction of the X-ray beam 3. The depth of the bank of leaves 31a is the greatest, the depth of the second bank 31b is less than the depth of the first bank of leaves and the depth of bank of leaves 31c is less than the depth of the second bank of leaves. The leaves are also curved in the direction parallel to the direction of the beam, with the radius of curvature of the first bank of leaves 31a being greater than the radius of curvature of the second bank of leaves 31b. The radius of curvature of the third bank of leaves 31c, is smaller than the radius of curvature of the second bank of leaves. FIG. 9 shows the resulting penumbra for this second aspect of the invention, which is improved almost two-fold compared with the conventional two-bank collimator. The percentage of the area outside of the collimator pattern that receives between 20–100% of the radiation for a conventional two-bank collimator is typically of the order of 13%. For the mini-collimator this is reduced to around 8%, but the multi-bank collimator according to the present invention, achieves a fluency of around 7%. Therefore, it can be seen that the multi-bank multi-leaf collimator of the present invention includes the advantages of the large two-bank conventional collimator in terms of the field size and yet has the fluency, resolution and performance of the mini-collimator. In addition, the use of a collimator according to the present invention allows the design of MLC heads that are significantly simpler. The leaves need not employ tongue and groove mechanisms at their joint, as other banks of leaves above and/or below can cover these joins. Furthermore, the collimator head does not need to be rotateable as the collimator is less directional. It will of course be understood that many variations may be made to the above-described embodiment without departing from the scope of the present invention.
	summary
	claims	1. In a microlithography method utilizing a charged-particle illumination beam to irradiate a portion of a pattern defined by a reticle situated on a reticle plane and a projection-optical system to direct a corresponding charged-particle imaging beam from the irradiated portion to a sensitized substrate situated on a substrate plane, an improved beam-alignment or calibration method comprising the steps: (a) defining at least one upstream mark on the reticle plane and at least one downstream mark on the substrate plane, each upstream mark being selectively registrable with a downstream mark; (b) providing a shield upstream of an upstream mark, the shield (i) serving to block downstream propagation of the illumination beam, and (ii) defining an aperture having a size and profile sufficient to pass therethrough only a portion of the illumination beam sufficient to irradiate the upstream mark; and (c) when irradiating an upstream mark with the illumination beam, passing the illumination beam through the aperture of the shield before the illumination beam reaches the upstream mark. 2. The method of claim 1 , wherein: claim 1 step (a) comprises defining at least one upstream mark on the reticle; and step (b) comprises extending the shield over the reticle. 3. The method of claim 2 , wherein: claim 2 step (a) comprises defining at least one upstream mark on a mark member; and step (b) comprises extending the shield over the mark member.
	summary
	abstract	The ion-implanting apparatus includes an implanting control device 26a having the functions of sweeping an ion beam by a sweeping magnet 12 and scanning a target by a scan mechanism. The implanting control device 26a has the functions of changing a sweep frequency of the ion beam to be swept by said sweeping magnet according to at least one of the species and energy of the ion beam and changing the minimum number of times of scanning of the target to be scanned by said scan mechanism according to the changing of the sweep frequency.
	summary
049884751	claims	1. Device for checking the axial retention force on a peripheral rod (6a) of a fuel assembly (1) of a nuclear reactor inside the framework of the assembly, comprising a plurality of spacer grids (2) retaining the fuel rods in a uniform network in transverse direction and in the axial direction of the rod (6a) by virtue of gripping means (15, 16, 7) associated with the cells of the grid (2) in which the rods (6) are inserted, the checking being performed remotely and under a certain depth of water in a fuel assembly storage pool, by virtue of the device which comprises: (a) a rod (28) on which is mounted a means (30, 32, 33) for support and displacement which is movable in an axial direction of the rod (28) and in two directions perpendicular to said axial direction; (b) a device for pushing axially on a longitudinal end of the rod (6a) carried by the means (30, 32, 33) for support and displacement, consisting of a fork (64) comprising an end notch (66) for its engagement on a shoulder of an end plug of the rod, said fork being fixed on an element (61) which is movable in the vertical direction in a guide means (60) fixed on the means (30, 32, 33) for support and displacement, and of at least one support fork on a face of a corresponding end joining piece (4, 5) of the fuel assembly, which fork is fixed on the means (30, 32, 33) for support and displacement, the vertical displacement of the movable element being ensured by a remote control means (42, 45); (c) means (77, 79, 76, 78) for measuring the axial pushing force on the rod (6a) and the amplitude of the axial displacement of the rod under the effect of the pushing action, disposed at the upper level of the storage pool (27); (d) and at least one video camera (34) carried by the means (30, 32, 33) for support and displacement in order to provide an image of the zone in the vicinity of the end of the rod (6a) on which a pushing action is being exerted; wherein the movable element consists of a sliding shaft (61) on which is machined a rack (61') of vertical direction, the remote control means (42, 45) consisting of a ball remote control whose flexible movable element (45) is connected to the end of a rack (47) engaging with a drive pinion (52) driving the displacement of the rack (61') and of the movable element (61) by means of at least one pinion (56, 57). 2. Device according to claim 1, wherein the end of the ball remote control (42, 45) opposite to its end which is connected to the rack (47) comprises an end of a sheath (42) fixed on a casing (71) and an end of a flexible element (45) which is movable in the sheath (42) fixed on a carriage (108) mounted so as to move inside the casing (71), an indicator (76) for measurement of displacement being fixed on the carriage (108) so as to comprise a movable end inside an aperture (75) passing through the casing (71) along a graduated rule (78) fixed on the outer surface of the casing (71) the indicator (76) and the graduated rule (78) making it possible to measure the displacements of the carriage (108) and of the movable flexible element (45) of the ball remote control. 3. Device according to claim 2, comprising means for displacement of the carriage (108) inside the casing (71), said means consisting of a screw (106) on which is engaged an internally threaded nut (107) interacting with the carriage (108) for its displacement guided by at least one column (101) fixed in the casing (71). 4. Device according to claim 3, wherein the screw (106) is mounted so as to move in rotation on a movable assembly (100) inside the casing (71) in the axial direction of the screw (106). 5. Device according to claim 4, wherein the movable assembly (100) comprises a sleeve (121), two end plates (104, 124, 105, 125) and a displacement device consisting of a threaded shaft (115) parallel to the screw (106) mounted so as to move in rotation in the casing (71) disposed in the bore of the sleeve (121) and along its axial direction, on which is mounted a threaded nut (118) engaging with the screw thread of the shaft (115) inside the sleeve (121), two dynamometric springs (127, 128) each being inserted between a support dish (119, 120) of the nut (118) and an end plate (104, 124, 105, 125) of the movable assembly (100), the threaded shaft (115) comprising a profiled end part outside the casing (71) for rotation of the shaft in one direction or another and for the axial displacement of the nut (118) inside the sleeve (121) ensuring the compression of a dynamometric spring (127, 128) leading to a displacement of the movable assembly (100) for a certain compression force corresponding to the pushing force necessary to overcome the axial retention force on the rod (6a). 6. Device according to claim 5, wherein the nut (118) carries an indicator (77) which is capable of being displaced inside an aperture (130) passing through the wall of the sleeve (121) coinciding with an aperture (74) passing through the casing (71) along a graduated rule (79) carrying a graduation indicating the pushing force on the pencil which is a function of the position of the nut (118). 7. Device according to claim 6, wherein the movable element (100) is guided in its displacements inside the casing (71) by two columns (101, 101') parallel to the screw (106) and to the threaded shaft (115) on which the movable assembly (100) is mounted so as to slide by means of ball bushes (103, 103') fixed on the end flanges (104, 105).
052232093	claims	1. A method for pressure relief of a containment of a nuclear power plant, which comprises: heating a washing fluid in a filter disposed inside a containment at a rated heating power through a thermal bridge, with a gas-steam mixture filling the containment, prior to initial operation of the filter; and rendering the thermal bridge substantially ineffective in an operating state of the filter, leaving the washing fluid with a continuous rated heating power being negligible in terms of filtration. sealing a filter disposed inside a containment against a gas-steam mixture filling the containment, prior to initial operation of the filter; heating a washing fluid in the filter through a thermal bridge with the gas-steam mixture at a given heating power, prior to the initial operation of the filter; allowing the gas-steam mixture to flow into the washing fluid for filtering therewith, in an operating state of the filter; and rendering the thermal bridge substantially ineffective in the operating state of the filter by reducing a heat transfer through the thermal bridge to the washing fluid, setting a continuous heating power being negligible in terms of filtration by the washing fluid. 2. The method according to claim 1, which comprises setting the continuous rated heating power at less than 0.1 times the rated heating power. 3. The method according to claim 1, which comprises raising the washing fluid to an operating temperature after at most eight hours of heating at the rated heating power. 4. The method according to claim 1, which comprises raising the washing fluid to an operating temperature after substantially two hours of heating at the rated heating power. 5. The method according to claim 1, which comprises setting an operating temperature in a range of substantially from 100.degree. to 150.degree. C., raising the operating temperature to substantially 260.degree. C. upon pressure relief directly from a primary loop of a nuclear power plant, and supplying water to the filter as the washing fluid. 6. The method according to claim 1, which comprises supplying an amount of heat with the continuous rated heating power being less than an amount of heat removed from the washing fluid by evaporation, for setting an operating temperature of the filter lower than an entry temperature of the gas-steam mixture. 7. A method for pressure relief of a containment of a nuclear power plant, which comprises: 8. The method according to claim 7, which comprises setting the continuous heating power at less than 0.1 times the given heating power.
	summary
046831088	description	DESCRIPTION OF THE PREFERRED EMBODIMENT Referring to FIG. 1, there is generally designated by the numeral 10 the core barrel of a nuclear reactor. Typically, there will be disposed within the core barrel 10 a fuel core assembly (not shown) of known construction, which is confined between upper and lower core plates. The core assembly is peripherally enclosed within a baffle comprising a zig-zag array of a plurality of baffle plates 11, with each baffle plate 11 being disposed substantially perpendicular to adjacent baffle plates 11. The space between the baffle plates 11 and the core barrel 10 is closed by a plurality of vertically spaced-apart former plates 12 which are seated in recesses 13 in the outer surfaces of the baffle plates 11 (see FIGS. 4 and 5). The baffle plates 11 are fixedly secured to the former plates 12 by a plurality of screw assemblies 15. More specifically, referring to FIGS. 2, 4 and 5, each baffle plate 11 has a plurality of screw bores 16 formed therethrough, each having an enlarged-diameter counterbore portion 17 at the inner end thereof which defines a recessed annular shoulder 18. Each of the screw bores 16 is disposed in axial alignment with a counterbore portion 19a of an internally threaded bore 19 in the associated former plate 12. In the original construction of the nuclear reactor, a suitable screw (not shown) is inserted in the bore 16 and threadedly engaged in the bore 19, the screw having a screw head which is recessed in the counterbore portion 17 and there lock welded to the baffle plate 11. Referring now in particular to FIGS. 2-6, there is disclosed a locking bolt apparatus 20 in accordance with the present invention, the apparatus 20 including a detent cutting tool 22 (FIG. 2), a screw 30, a locking cup 35 and a staking tool assembly 40. The screw 30 and the locking cup 35 cooperate, after operation thereon by the staking tool assembly 40, to provide a locking screw assembly 60 in accordance with the present invention. The detent cutting tool 22 is provided at its distal end with a cutting tip 24 carrying suitable cutting members 24a. The cutting tool 22 is adapted to be rotatably mounted on a suitable support and drive mechanism (not shown) which can be lowered into the core barrel 10 for inserting the cutting tip 24 axially into the counterbore portion 17 of a screw bore 16 which has had a damaged screw removed therefrom. The cutting tool 22 is then moved radially and rotated to machine a part-spherical, radially outwardly extending lateral recess 25 in the counterbore portion 17. Any desired number of such recesses 25 may be formed. In the preferred embodiment four equiangularly spaced-apart recesses 25 are machined in the counterbore portion 17. When the machining operation is completed the screw 30 can be inserted in the screw bore 16. The screw 30 has an elongated shank 31 provided at one end with an enlarged-diameter externally threaded portion 32. The other end of the shank 31 is provided with a radially outwardly extending annular shoulder flange 33 which preferably has a diameter slightly less than that of the counterbore portion 17 of the screw bore 16. Integral with the shoulder flange 33 and projecting axially therefrom is a hexagonal drive head 34 which has a lateral width substantially less than that of the shoulder flange 33. The locking cup 35 has a circular end wall 36 provided with a hexagonal opening 37 formed centrally therethrough. Integral with the circular end wall 36 around the outer periphery thereof and projecting therefrom coaxially therewith is a cylindrical side wall 38. Preferably, the cylindrical side wall 38 has a thickness substantially less than that of the circular end wall 36 so as to facilitate deformation thereof, as will be explained more fully below. The hexagonal opening 37 is dimensioned to receive the drive head 34 therethrough in close fitting relationship, i.e., in a frictional or slight interference fit, with the circular end wall 36 disposed against the shoulder flange 33. Thus, if desired, the locking cup 35 can be preassembled with the screw 30 before the screw 30 is mounted in the screw bore 16. The screw 30 is inserted into the screw bore 16 and threadedly engaged in the bore 19 until the shoulder flange 33 seats firmly against the shoulder 18 of the counterbore portion 17 with whatever torque value is required, as illustrated in FIG. 4. When the parts are thus assembled, the locking cup 35 is completely recessed in the counterbore portion 17, but the cylindrical side wall 38 extends upwardly a distance sufficient to completely cover the lateral recesses 25. The staking tool assembly 40 includes an elongated cylindrical die block 41 having a generally cylindrical bore 42 extending axially therethrough. Formed in one end of the die block 41 is a diametrical slot 43 which bifurcates the die block 41 and separates it into a pair of legs 44. Each of the legs 44 is provided with a flat part-circular bearing surface 45. Integral with the bearing surface 45 and projecting downwardly therefrom on each of the legs 44 is a die finger 46 shaped complementary to the recesses 25, each of the die fingers 46 having a downwardly and inwardly tapered inner cam surface 47 (see FIG. 5). The staking tool assembly 40 also includes an elongated cylindrical drive bolt 50 dimensioned to slidably fit in the bore 42. The drive bolt 50 has a frustonconical tapered drive end 51 in which is formed an axial bore or recess 52. In operation, the staking tool assembly 40 is carried by a suitable drive mechanism (not shown). The die block 41 is seated against the surface of the baffle plate 11 in surrounding relationship with the counterbore portion 17 of the screw bore 16, and with the die fingers 46 extending downwardly into the counterbore portion 17 and into the locking cup 35 in positions respectively opposite the lateral recesses 25. To facilitate accurate rotational positioning of the die block 41, suitable positioning marks may be provided on the surface of the baffle plate 11. When the die block 41 has been thus positioned, the drive belt 50 is then driven axially in the direction of the arrow in FIG. 5 for driving the tapered drive end 51 into camming engagement with the cam surfaces 47 and laterally deflecting the die fingers 46 radially outwardly against the cylindrical side wall 38 of the locking cup 35, thereby deforming the side wall 38 into the recesses 25 to form a plurality of locking projections 55 (see FIG. 6) which are engaged firmly in the recesses 25. This lateral deflection of the die fingers 46 is facilitated by the bifurcation of the die block 41. The staking tool assembly 40 is then removed and there results a completed locking assembly 60 (FIG. 6). The screw 30 is firmly threadedly engaged with the former plate 12. The locking cup 35 cooperates with the drive head 34 to prevent relative rotation of those two parts, while the engagement of the projections 55 in the lateral recesses 25 prevents any movement of the locking cup 35 with respect to the baffle plate 11. Thus, it will be appreciated that the screw 30 is securely retained against loosening rotation and, furthermore, the locking cup 35 acts securely to trap the screw drive head 34 and shoulder flange 33 to prevent accidental dislodgement thereof in the event of breakage of the shank 31, thereby resulting in an effective Class A locking device. It is a significant aspect of the invention that the recesses 25 are disposed beneath the surface of the baffle plate 11, thereby avoiding machining of that surface which can be embrittled due to reactor core exposure. Furthermore, in the event that an error is made in the machining of one of the recesses 25, the error is easily corrected since the entire circumferential surface of the counterbore portion 17 remains for the formation of other recesses 25. From the foregoing, it can been seen that there has been provided an improved locking screw assembly and apparatus and method for installation thereof, which is of relatively simple and economical construction, is suitable for remote underwater application, results in a Class A locking device, is suitable for use in replacing the baffle plate bolts of a nuclear reactor core barrel without machining the exposed surface of the baffle plate, and which involves relatively low risk of damage to the baffle plate or irreparable error in the installation operation.
	description	1. Field of the Invention The present invention generally relates to exchanging packets of data on an interconnect bus connecting two devices, and more particularly, to measuring and verifying the performance of such an exchange. 2. Description of the Related Art A system on a chip (SOC) generally includes one or more integrated processor cores, some type of embedded memory such as a cache shared between the processor cores, and peripheral interfaces such as an external bus interface, on a single chip to form a complete (or nearly complete) system. The external bus interface is often used to pass data in packets over an external bus between these systems and an external device such as an external memory controller, Input/Output (I/O) controller, or graphics processing unit (GPU). The performance of such a system may depend on several factors which may include device characteristics, characteristics of interconnect buses, memory hierarchy, operating system, and various other factors. A reasonable prediction of ranges for system performance can still be made after considering such factors. However, it is generally desirable to verify that performance falls within these ranges during simulation. For example, it may be desirable to verify that the throughput (or bandwidth) and the latency (or response time) of communication over an interconnect bus between a transmitting and receiving device fall within their predicted range. Conventionally, simulation involves running predefined test cases modeled to emulate normal system operation. During simulation, bus traffic is monitored, interesting events on the bus are captured, and performance is measured based on the captured events. The captured events and their performance metrics are recorded in a simulation log. It is only after simulation that a user can view all the bus events in the simulation log and identify categories of events that fall outside the predicted performance range. However, because the information contained in the simulation logs is rather cryptic, significant effort will be required to manually analyze, identify and parse those categories of events that do not fall within their performance range. Another problem with conventional simulation is that predefined test cases may not adequately test a given category of bus events. For example, a test case may not contain a sufficient number of read operations. As a result, the performance measurements for the read operation may not be statistically significant. Yet another problem with the conventional testing method is that degradations in performance are unlikely to be detected, without tedious manual analysis, when the predicted range of performance is too lenient. For example, if the average latency associated with a particular transaction between two devices is predicted to be 1 second, but the measured average latency is only 0.2 seconds, then a degradation of the average latency from 0.2 seconds to 0.8 seconds is unlikely to be caught even though there is a significant, undesired change in performance. Accordingly, what is needed is improved methods and apparatus for measuring and verifying performance of packet based data exchanges between devices connected by an interconnect bus. Embodiments of the present invention generally provide methods, computer readable storage media, and systems for measuring and verifying performance of packet based communication transactions between devices over an interconnect bus. One embodiment provides a method for determining performance characteristics of a system. The method generally includes executing a program to cause data to be exchanged between at least two devices of the system via a bus, capturing events indicative of data exchanged between the at least two devices by at least one interface monitor, calculating one or more performance metrics based on the captured events during execution of the program, storing the calculated performance metrics in a database, and determining whether the calculated performance metrics fall within a determined performance range. Another embodiment provides a computer readable storage medium containing a program for determining performance characteristics of a system. When executed by a processor, the program performs operations generally including generating data to be exchanged between at least two devices of the system via a bus, capturing events indicative of data exchanged between the at least two devices by at least one interface monitor, calculating one or more performance metrics based on the captured events during execution of the program, storing the calculated performance metrics in a database, and determining whether the calculated performance metrics fall within a determined performance range. Yet another embodiment provides a system generally including a first processing device, a second processing device coupled with the first processing device via a bus, at least one interface monitor for capturing events indicative of data exchanged between the at least two processing devices via the bus, and a performance monitor configured to calculate one or more performance metrics based on the captured events, store the one or more calculated performance metrics in a database, and determine whether the calculated performance metrics fall within a determined performance range. Embodiments of the present invention allow packet based communication transactions between devices over an interconnect bus to be captured to measure performance. Performance metrics may be determined by capturing events at various nodes as they pass through the system. Performance may be verified at run time by computing performance metrics for captured events and comparing such metrics to predefined performance ranges and/or self learned performance ranges. Furthermore, embodiments of the present invention provide for dynamic tailoring of bus traffic to generate potential failing conditions. In the following, reference is made to embodiments of the invention. However, it should be understood that the invention is not limited to specific described embodiments. Instead, any combination of the following features and elements, whether related to different embodiments or not, is contemplated to implement and practice the invention. Furthermore, in various embodiments the invention provides numerous advantages over the prior art. However, although embodiments of the invention may achieve advantages over other possible solutions and/or over the prior art, whether or not a particular advantage is achieved by a given embodiment is not limiting of the invention. Thus, the following aspects, features, embodiments and advantages are merely illustrative and not considered elements or limitations of the appended claims except where explicitly recited in the claim(s). Likewise, reference to “the invention” shall not be construed as a generalization of any inventive subject matter disclosed herein and shall not be considered to be an element or limitation of the appended claims except where explicitly recited in a claim(s). FIG. 1 illustrates an exemplary testing system in which a Performance Monitor 100 monitors performance between two devices (or nodes) 120 and 130 over an Interconnect Bus 180 (e.g., commonly referred to as a front side bus). The two devices 120 and 130, for example, may be a central processing unit (CPU) and a graphics processing unit (GPU). For some embodiments, the Bus 180 may be a bi-directional multi-bit bus, for example, having eight or more lines for communication from the CPU to the GPU and another eight or more lines for communication from the GPU to the CPU. Communication between the devices 120 and 130 may be monitored by a Link Interface Monitor (IM) 140. Link IM 140 may be any combination of hardware and/or software configured to sample data lines of the Interconnect Bus 180 in conjunction with a clock signal. The Link IM may be further configured to examine the sampled data and recognize predefined categories of events. If a known event is captured, the Link IM may notify the performance monitor that the event is presented on the Interconnect Bus 180. For example, a CPU may perform a read operation on a specific location in the GPU by sending a read packet over an interconnect bus connecting the CPU and the GPU. The Link IM for the interconnect bus may capture the read packet when it is presented on the bus and notify the performance monitor that a read packet is found on the bus. In some embodiments of the invention the Link IM may be configured to inject noise on to the Interconnect Bus 180. Such noise injection may be performed to simulate actual noise on the interconnect bus during normal operation of the system. In other embodiments, the Link IM may also be configured to introduce errors into an event captured on the bus before the event is dispatched to the destination device. For example, the Link IM may toggle some bits in the packet. As with noise injection, the introduction of errors may be performed to simulate actual errors that may occur while transferring packets during normal operation of the system. The goal of introducing such errors may be to verify that the destination device properly determines an error in the packet, for example by using a Cyclic Redundancy Check (CRC), and performs error correcting steps which may include correcting erroneous bits or requesting that the packet be sent again. While the above mentioned embodiments describe noise and error injection performed by the Link IM, those skilled in the art will recognize that such noise and error injection may be performed by a separate and independent device, such as an irritator device. Each device 120 and 130 may be driven by Unit Drivers 160 and 161 respectively. Each Unit Driver may be software that is configured to cause an associated device to perform a series of functions, including sending packets to another device. For example, Unit Driver 160 may generate instructions to Device 120 to send a read packet to Device 130 over the Interconnect Bus 180. Such instructions by unit drivers 160 and 161 to devices 120 and 130 may be monitored by Application Interface Monitors (IM) 150 and 151 respectively. Each Application IM may be any combination of hardware and/or software configured to sample data lines connecting the Application IM and an associated device in conjunction with a clock signal. Furthermore, each Application IM may be configured to examine the sampled data and recognize predefined categories of instructions. As with the Link IM, if a known instruction is captured, the Application IM may notify the performance monitor that the event is presented to the associated device. The events captured by the Link IM 140 and the Application IM's 150 and 151 may be received by a Performance Monitor 100 and stored in a shared Database 170. In some embodiments of the invention the Performance Monitor may store in the Database 170, a timestamp associated with each captured event. For example, the Performance may store in the database the simulation time at which each event was captured by the interface monitors. The Performance Monitor 100 may be configured to calculate several performance metrics for the system based on the captured events. For example, to compute the latency of a read operation across Device 120, the Performance Monitor may subtract the timestamps for a read instruction issued by Unit Driver 160 and captured by Application IM 150, and an associated read packet captured by Link IM 140. Similarly, the Performance Monitor may also compute the latency of read responses between Device 1 and Device 2 over the Interconnect Bus 180 by subtracting the timestamps of a read packet and an associated data packet captured by Link IM 140. Several other similar performance metrics may be defined to measure latencies and throughput for the system. The Performance Monitor may be further configured to store the calculated performance metrics in the shared Database 170. For example, the Performance Monitor may store the latencies of write and read operations in Database 170. A user may be allowed to query Database 170 to generate graphs that illustrate performance results for various bus events. Such graphs may allow a user to easily compare results between bus events in the same test run and/or different test runs. The Performance Monitor 100 may be configured to fail a simulation test based on predefined or self learned performance ranges 101. The performance ranges 101 may define upper and lower range limits or an upper or lower threshold value. A predefined range may be defined by a user before running a test on the system. The predefined ranges may be chosen arbitrarily or according to ideal performance metrics calculated considering factors such as device characteristics, system architecture, system software, and the like. The self learned ranges, on the other hand, may be calculated based on historic system performance data contained in Database 170. For example, the self-learned ranges may be determined by computing an average of previously obtained performance metrics or by selecting values at or near the peak of a bell curve representing historic performance results. Any other reasonable method for calculating performance ranges may be used to determine expected performance based on historic performance. FIG. 2 is a flow diagram for exemplary operations performed to capture and store bus events in accordance with embodiments of the present invention. The operations may be performed, for example, by components illustrated in FIG. 1, while executing a specific program designed to emulate normal system operation (and produce typical bus traffic). However, those skilled in the art will recognize that the operations of FIG. 2 may be performed by other components and, further, that the components illustrated in FIG. 1 may be capable of performing other operations. The operations begin, at step 201, by capturing events on the bus. As previously described, a Link IM or an Application IM may detect events indicating a transaction between devices or between a unit driver and an associated device, capture such an event, and send it to a Performance Monitor. In some embodiments of the invention the Link IM and Application IM may be a part of the Performance Monitor, therefore the events may be captured by the performance monitor directly. Captured events may be stored in a shared Database 204, as illustrated. At step 202, the Performance Monitor may interpret the captured event and calculate Performance metrics for that event. This may require the Performance Monitor to query the database to find other events associated with the captured event. For example, when the Performance Monitor captures a read packet on the Interconnect Bus 180, it may query Database 170 for a read instruction issued from the Unit Driver 160 in order to calculate the latency of the read operation through Device 120. Several other performance metrics may also be computed at this time. At step 203, the Performance Monitor may store the calculated performance metrics in the shared database. The performance metrics stored in the database may be used later to compute self learned ranges for system performance. FIG. 3 is a flow diagram for exemplary operations performed to verify, during run time, that performance of a system falls within predefined ranges. The operations begin at step 301 by getting the user defined ranges. At this step, the user may be prompted to define ranges for one or more performance metrics. Alternatively, the user may also be allowed to select predefined ranges used in previous simulations. Sets of predefined ranges may also be organized into test profiles. Each test profile may contain a unique combination of performance range settings. A user may be prompted to select one of these profiles at the outset of simulation. In one embodiment of the invention, the predefined ranges may be selected for a plurality of simulation tests to facilitate batch testing with the same predefined parameters. At step 302, simulation begins by Unit Drivers generating stimulus to the devices in order to emulate normal system operation and produce typical bus traffic. As simulation continues, the Performance Monitor performs the steps outlined in FIG. 2 to capture events and measure performance. In some embodiments of the invention, the Performance Monitor may compute performance results only after the simulation is run for a predetermined period of time. As each event is captured and performance metrics calculated, the test in step 303 is performed to determine whether the performance metrics calculated fall within the predefined ranges. If a calculated performance metric for a captured event falls outside of its predefined range, simulation may be stopped and a system failure message may be generated at step 306. In some embodiments of the invention, simulation may be stopped only if a certain threshold number of events fall outside the predefined range. Stopping simulation on the occurrence of a failing condition may save valuable simulation time and make performance verification more efficient. If, on the other hand, the performance metric is deemed to fall within the predefined range, the Performance Monitor continues to capture and calculate performance metrics for events until another performance metric falls outside the predefined range or an end-of-test is detected in step 304. If an end-of-test is detected and all performance metrics fall within the predefined ranges, then the simulation run is deemed successful at step 305. FIG. 4 is a flow diagram for exemplary operations performed to verify, during run time, that performance of a system falls within ranges determined by the system (self learned ranges). The operations begin in step 401 by determining the ranges that will be used to verify performance metrics. The ranges may be determined by querying the Database 170 for performance metrics stored from previously run simulations and computing the self learned ranges based on such historic data. As discussed earlier, any method such as computing averages and normal curve peaks may be used to determine an expected performance range based on historic data. In step 402, the simulation may begin once the self learned performance ranges are determined. As in the description for FIG. 3, the Performance Monitor may monitor and calculate the performance metrics for events as they are captured during run time. These calculated performance metrics may be stored for later calculations of self learned ranges. In some embodiments of the invention, however, the Performance monitor may use the calculated performance metric for a captured event to dynamically update the self learned ranges being applied in the current simulation. In step 403, if a calculated performance metric for a captured event falls outside of the self learned range, simulation may be stopped and a system failure message may be generated at step 406. In some embodiments of the invention, simulation may be stopped only if a certain threshold number of events fall outside the self learned range. If, on the other hand, the performance metric is deemed to fall within the predefined range, the Performance Monitor continues to capture and calculate performance metrics for events until another performance falls outside the self learned range or an end-of-test is detected in step 404. If an end-of-test is detected and all performance metrics fall within the predefined ranges, then the simulation run is deemed successful at step 405. In some embodiments of the invention, the user may be allowed to configure the Performance Monitor to compare the performance metrics for a captured event with predefined ranges, self learned ranges, or both the predefined ranges and self learned ranges. For example, a user may choose to run simulation according to user defined ranges when the Database 170 does not contain sufficient information to calculate statistically significant self learned ranges. On the other hand, a user may run simulation according to the self learned ranges in order to detect any drastic changes in performance when the predefined ranges are suspected to be too lenient. Alternatively, a user may elect to run simulation according to both the predefined and self learned ranges to obtain the benefits of both approaches to verifying performance. One common problem with using predefined test cases to generate traffic during simulation is that a problem causing event may not be adequately tested by the test case. For example, a test case may have only a few read operations which may be insufficient to bring about a failing condition. Therefore, another test case must be written that has sufficient read operations. However, under this approach an innumerable number of test cases will have to be written to account for all the various permutations and combinations of failing conditions. The present invention provides for dynamically tailoring the events generated by the Unit Drivers by weighting commands based on run time results. For example, if a write operation latency is deemed to be approaching a failing condition, a weight parameter associated with the write operation may be dynamically adjusted so that the write operation is generated more frequently. One method for determining whether a performance metric is approaching a failing condition may be to determine if the performance metric falls outside a threshold range within the predefined range and/or self-learned range. Referring back to FIG. 1, the Performance Monitor 100 may contain the necessary logic to compute weights for different categories of events based on run time results and provide feedback to the Unit Drivers 160 and 161. In response to this feedback, the Unit Drivers may dispatch instructions to reflect the dynamically adjusted weights for the instructions. By monitoring key performance metrics real time during simulation, then using that information along with predefined and/or self learned performance ranges and dynamic command weighting based on real time results to fail the simulation, the present invention may notify a user that there is a potential problem, and identify the offending event. As a result, a more efficient and effective verification of system performance may be achieved. While the foregoing is directed to embodiments of the present invention, other and further embodiments of the invention may be devised without departing from the basic scope thereof, and the scope thereof is determined by the claims that follow.
	claims	1. A method for forming a radiation flood source, the method comprising:preparing a radioactive isotope carrier solution;loading the radioactive isotope carrier solution into a radioactive isotope carrier solution cartridge;loading a separate border cartridge into a plotter;selecting and configuring a shape of an active area;setting a border to be placed around the active area;printing the active area by utilizing the radioactive isotope carrier solution cartridge on a sheet substrate; andprinting the border by utilizing the separate border cartridge on the sheet substrate. 2. The method of claim 1, further comprising:laminating the printed sheet substrate to block radioactive isotopes on the active area from separating from the active area; andplacing the laminated printed sheet substrate in a protective housing. 3. The method of claim 1, further comprising:cutting an active sheet around the printed border from the printed sheet substrate;laminating the active sheet to block radioactive isotopes on the active area from separating from the active area; andplacing the laminated active sheet in a protective housing. 4. The method of claim 1, further comprising:cutting an active sheet around the printed border from the printed sheet substrate;laminating the active sheet to block radioactive isotopes on the active area from separating from the active area;testing the laminated active sheet to verify an integrity of the active area; andplacing the tested active sheet in a protective housing. 5. The method of claim 1, wherein the preparing of the radioactive isotope carrier solution comprises:drying a radioactive isotope solution to form dried radioactive isotopes; andmixing the dried radioactive isotopes with a pigmentless carrier solution to prepare the radioactive isotope carrier solution. 6. The method of claim 1, wherein the active area is printed only by the radioactive isotope carrier cartridge, and the border is printed only by the border cartridge. 7. The method of claim 1, wherein the radioactive isotope carrier solution comprises an active material composed of radioactive isotopes selected from the group consisting of Cobalt 57, Iodine 125, Palladium 103, Barium 133, Carbon 14, Gadolinium 153, Phosphorus 33, Tellurium 99, and combinations thereof. 8. The method of claim 7, wherein the radioactive isotope carrier solution is formulated with a pigmentless carrier solution comprising cobalt chloride, ethylene glycol, glycerin, and hydrochloric acid and to have a viscosity adapted for being inkjet printed on the sheet substrate. 9. The method of claim 8, wherein the pigmentless carrier solution is composed of a mixture of 600 mg of cobalt chloride, 10 ml ethylene glycol, 10 ml glycerin, and 80 ml of 0.1M hydrochloric acid. 10. The method of claim 1, wherein:the radioactive isotope carrier solution is a pigmentless radioactive isotope carrier solution;the printing of the active area comprises printing the active area by utilizing only the pigmentless radioactive isotope carrier solution;the separate border cartridge comprises a pigmented ink solution; andthe printing of the border comprises printing the border around the area by utilizing only the pigmented ink solution. 11. A plotting system for forming a radiation flood source comprising:a sheet substrate supply configured to provide a sheet substrate;a radioactive isotope carrier solution cartridge containing a radioactive isotope carrier solution and configured to print an active area onto the sheet substrate;a separate border cartridge configured to print a border around the active area on the sheet substrate; anda controller configured to control the radioactive isotope carrier solution cartridge to print the active area onto the sheet substrate and the separate border cartridge to print the border around the active area on the sheet substrate. 12. The plotting system of claim 11, wherein the radioactive isotope carrier solution cartridge is an inkjet cartridge. 13. The plotting system of claim 11, wherein the radioactive isotope carrier solution is a mixture of dried radioactive isotopes and a pigmentless carrier solution. 14. The plotting system of claim 11, wherein the active area is printed only by the radioactive isotope carrier cartridge, and the border is printed only by the border cartridge. 15. The plotting system of claim 11, wherein the radioactive isotope carrier solution comprises an active material composed of radioactive isotopes selected from the group consisting of Cobalt 57, Iodine 125, Palladium 103, Barium 133, Carbon 14, Gadolinium 153, Phosphorus 33, Tellurium 99, and combinations thereof. 16. The plotting system of claim 11, wherein the radioactive isotope carrier solution is formulated with a pigmentless carrier solution comprising cobalt chloride, ethylene glycol, glycerin, and hydrochloric acid and to have a viscosity adapted for being inkjet printed on the sheet substrate. 17. The plotting system of claim 16, wherein the pigmentless carrier solution is composed of a mixture of 600 mg of cobalt chloride, 10 ml ethylene glycol, 10 ml glycerin, and 80 ml of 0.1M hydrochloric acid. 18. The plotting system of claim 11, wherein the separate border cartridge contains a pigmented solution composed of color pigments selected from the group consisting of black pigments, cyan pigments, yellow pigments, magenta pigments, and combinations thereof. 19. A radiation flood source comprising:a paper sheet;a pigmentless radioactive fill printed on the paper sheet and comprising radioactive isotopes selected from the group consisting of Cobalt 57, Iodine 125, Palladium 103, Barium 133, Carbon 14, Gadolinium 153, Phosphorus 33, Tellurium 99, and combinations thereof; anda pigmented border printed on the paper sheet and around the pigmentless radioactive fill. 20. The radiation flood source of claim 19, further comprising:a first protective sheet laminated with the paper sheet with the radioactive isotopes therebetween. 21. The radiation flood source of claim 20, further comprising a second protective sheet and the paper sheet being laminated between the first protective sheet and the second protective sheet. 22. The radiation flood source of claim 21, further comprising a housing having an interior space housing the paper sheet with the pigmentless radioactive fill. 23. The radiation flood source of claim 22, further comprising a spacer also housed in the interior space of the housing and between an interior side of the housing facing the paper sheet and the paper sheet. 24. The radiation flood source of claim 19, further comprising a housing having an interior space housing the paper sheet with the pigmentless radioactive fill. 25. The radiation flood source of claim 24, further comprising a spacer also housed in the interior space of the housing and between an interior side of the housing facing the paper sheet and the paper sheet. 26. The radiation flood source of claim 19, wherein the pigmentless radioactive fill further comprises a pigmentless carrier material. 27. The radiation flood source of claim 19, wherein the pigmented border comprises color pigments selected from the group consisting of black pigments, cyan pigments, yellow pigments, magenta pigments, and combinations thereof. 28. The radiation flood source of claim 19, wherein the pigmentless radioactive fill is transparent to visible light.
043671950	abstract	Apparatus for homogenizing the circumferential temperatures of the vertically axes ferrule of a component passing through the upper slab of a nuclear reactor, wherein it comprises at least one assembly forming a heat pipe describing the entire circumference of said ferrule in order to ensure the homogenization of the temperatures of said ferrule level with the assembly, means for fixing the assembly or assemblies forming the heat pipe on the inner face of the ferrule and means for ensuring a thermal contact between the assembly or assemblies forming the heat pipe and the said ferrule.
059149949	abstract	A storage basket, a storage rack and a method are provided for the compact storage of fuel elements and control rods of a nuclear power plant. The storage basket includes a plurality of inserts for receiving a fuel element. The inserts are disposed in such a way as to form a cruciform gap for receiving a control rod. Four fuel elements as well as one control rod of a boiling water nuclear power plant can be intermediately stored in a carrying well of a fuel element storage rack in a particularly compact and combined way by placing inserts in such a manner as to form a storage basket. A checkered configuration of the carrying wells of the fuel element storage rack permits intermediate positions between the carrying wells to also be supplied with one control rod and four fuel elements.
	claims	1. A dry radioactive substance storage facility comprising: a structure having a storage room storing storage vessels containing a radioactive substance; an air inlet duct having an air inlet and defining an air inlet passage through which air is supplied into the storage room; a stack having an air outlet and defining an air discharge passage through which air from the storage room is discharged outside; a plurality of first baffling members disposed between a ceiling portion of the storage room and a floor portion of the storage room on the side of the air inlet duct in the storage room and being adjacent to the storage vessel, and arranged so that air supplied through the air inlet duct flows through spaces between the first baffling members into the storage room; and a plurality of second baffling members disposed between a ceiling portion of the storage room and a floor portion of the storage room on the side of the stack in the storage room and being adjacent to the storage vessel, and arranged so that air in the storage room flows through spaces between the second baffling members into the stack; wherein each of the first and the second baffling members comprises a neutron shielding member and a gamma ray shielding member. 2. The dry radioactive substance storage facility according to claim 1 , wherein each of the first and the second baffling members is formed by covering the neutron shielding member with the gamma ray shielding member. claim 1 3. The dry radioactive substance storage facility according to claim 1 , wherein at least some of the second baffling members are arranged so that upper ones are shifted toward the stack relative to the lower ones. claim 1 4. The dry radioactive substance storage facility according to claim 2 , wherein at least some of the second baffling members are arranged so that upper ones are shifted toward the stack relative to the lower ones. claim 2 5. The dry radioactive substance storage facility according to claim 1 , wherein the first baffling members are inclined so as to direct air supplied through the air inlet passage toward an upper region of the storage room. claim 1 6. A dry radioactive substance storage facility comprising: a structure having a storage room storing storage vessels containing a radioactive substance; an air inlet duct having an air inlet and defining an air inlet passage through which air is supplied into the storage room; a stack having an air outlet and defining an air discharge passage through which air from the storage room is discharged outside; a plurality of first radiation shielding members disposed between a ceiling portion of the storage room and a floor portion of the storage room on the side of the air inlet duct in the storage room and being adjacent to the storage vessel; and a plurality of second radiation shielding members disposed between a ceiling portion of the storage room and a floor portion of the storage room on the side of the stack in the storage room and being adjacent to the storage vessel; wherein first air passages are formed between the first radiation shielding members, and second air passages are formed between the second radiation shielding members; first radiation scattering members are disposed on the side of the air inlet passage relative to the first radiation shielding members in the storage room so as to attenuate radiation propagating in the direction of the air inlet passage from the first radiation shielding members; and second radiation scattering members disposed on the side of the air discharge passage relative to the second radiation shielding members in the storage room so as to attenuate radiation propagating in the direction of the air discharge passage from the second radiation shielding members. 7. The dry radioactive substance storage facility according to claim 6 , wherein each of the first and second radiation scattering members comprises a neutron shielding member and a gamma ray shielding member. claim 6 8. The dry radioactive substance storage facility according to claim 7 , wherein each of the first and second radiation scattering members is formed by covering the neutron shielding member with the gamma ray shielding member. claim 7 9. The dry radioactive substance storage facility according to claim 6 , wherein a surface of each of the first and second radiation scattering members facing the storage room is formed in a shape capable of reflecting radiation downward. claim 6 10. The dry radioactive substance storage facility according to claim 1 , further comprising first radiation scattering members disposed on the side of the air inlet passage relative to the first baffling members in the storage room so as to attenuate radiation propagating in the direction of the air inlet passage from the first baffling members; and claim 1 second radiation scattering members disposed on the side of the air discharge passage relative to the second baffling members in the storage room so as to attenuate radiation propagating in the direction of the air discharge passage from the second baffling members. 11. The dry radioactive substance storage facility according to claim 10 , wherein each of the first and second radiation scattering members comprises a neutron shielding member and a gamma ray shielding member. claim 10 12. The dry radioactive substance storage facility according to claim 11 , wherein each of the first and second radiation scattering members is formed by covering the neutron shielding member with the gamma ray shielding member. claim 11 13. The dry radioactive substance storage facility according to claim 10 , wherein a surface of each of the first and second radiation scattering members facing the storage room is formed in a shape capable of reflecting radiation downward. claim 10
055419690	summary	BACKGROUND OF THE INVENTION The present invention is directed to a water level indicator for use in a nuclear power plant. During maintenance periods at a nuclear power plant, it is important for safety reasons to know the level of water in the hot leg of the water line travelling from the reactor vessel to the heat exchanger to ensure proper reactor core cooling. Present methods of measuring the water in this particular pipe have proven inadequate because of inaccuracies in the measurement of the water level in the pipe or the inability to make the water level reading from the control room of the power plant. During nuclear plant operation, water is heated under pressure in the reactor vessel. The water then travels through the hot leg pipe of the water circulation system to the heat exchanger/steam generator where the water is cooled. The cooled water then travels through a circulation pump to be returned to the reactor vessel for reheating. The water serves as the medium to cool the reactor core and transfer heat to the heat exchanger steam generator. During maintenance periods, water needs to be drained from the heat exchanger/steam generators to facilitate examination of the generator's internal equipment. While the steam generator does not serve its heat exchanging function during these maintenance periods, the water is cooled by another heat exchanger as the water continues to function as a coolant for the core. Since the hot leg pipe lies in essentially a horizontal plane, the precise level of water in the pipe is important. Should the water in the pipe be too high, or full, the inside of the generator, where work crews are performing examinations, could be flooded. Should the water level in the hot leg pipe be too low, or empty, this could indicate a water level too low to maintain proper cooling of the nuclear core. Because the hot leg has a direct fluid connection to the reactor vessel, the hot leg is in a convenient position for measuring the water level to ensure proper water level in the core. Therefore, being able to measure the precise water level in the hot leg pipe is of crucial importance for safety. It is of additional safety consideration that the level may be measured from the nuclear plant's control room where other safety equipment is monitored. SUMMARY OF THE INVENTION The primary object of the present invention is to provide apparatus to accurately measure the level of water in the hotleg of a pressurized water reactor system. It is a further object of the invention that the measurement of the level of the water can be monitored from a remote location such as the control room. It is another objective of the invention that the monitoring can be performed during all periods of plant operation and maintenance. It is still another objective of the invention that a measurement system can be permanently installed on the hot water leg pipe. It is yet another object of the invention to have a water level measurement system that can selectively be isolated from the hot leg water pipe. In fulfillment of these and other objectives, the midloop water level monitor of the present invention comprises a remote chamber fluidly connected to the pipe for measurement of the level of water in the pipe. In particular, a remote tank forming a chamber is located at an elevation substantially the same as that of the hot water pipe. A lower connecting pipe extends from the bottom of the hot water pipe to the bottom of the chamber and an upper connecting pipe extends from the top of the hot water pipe to the top of the chamber. The result is that the level of water in the chamber is substantially the same as the level of water in the pipe. Inside the chamber are means such as heated junction thermocouples for measuring the level of water in the chamber. It is also possible to use mechanical measurement systems, such as a float, connected to a rheostat or other electrical means of generating an electrical signal commensurate with the water levels. Isolation valves are located on each of the connecting pipes to allow isolation of the chamber from the main pipe such as during power generating operation of the power plant. This allows maintenance of the water level monitor without disrupting plant operation. In addition, because the water level monitor need only be at the same elevation as the hot leg pipe, the monitor can be located at any remote distance from the hot leg pipe. The isolation valves allow maintenance at this remote location during plant operation. These isolation valves can also be used to protect the detection equipment in the chamber from the pressure and heat of the water traveling through the pipe when the plant is in full operation. Since the hot water pipe lies in a substantially horizontal plane, it is important to be able to measure the water in this pipe with accuracy because a change of even a fraction of an inch within this pipe can result in an excessive level of water, which could flood the steam generator chamber during an examination by plant personnel, or an insufficient level of water, which could result in degradation of cooling of the reactor core .
052767256	claims	1. An exposure apparatus for exposing an original having a pattern and a substrate, with synchrotron radiation, said apparatus comprising: means for providing the synchrotron radiation; an exposure unit being able to support the original and the substrate, and for exposing the original and the substrate with the synchrotron radiation to transfer the pattern of the original onto the substrate, said exposure unit having a reflective member; a detector for receiving any of the synchrotron radiation as reflected by said reflective member of said exposure unit, to detect the attitude of said exposure unit with respect to the synchrotron radiation; and means for reducing distortion related to the transfer of the pattern, on the basis of the detection by said detector. means for providing the synchrotron radiation; an exposure unit capable of supporting the original and the substrate, for exposing the original and the substrate with the synchrotron radiation, said exposure unit including a reflective member; a detector for receiving any of the synchrotron radiation as reflected by said reflecting member of said exposure unit to detect the attitude of said exposure unit with respect to the synchrotron radiation; and an adjuster for adjusting the attitude of said exposure unit with respect to the synchrotron radiation, on the basis of the detection by said detector. introducing synchrotron radiation into an exposure unit, the exposure unit being capable of supporting an original and a substrate therein; detecting any of the synchrotron radiation as reflected in the exposure unit to detect the attitude of the exposure unit with respect to the synchrotron radiation; adjusting the attitude of the exposure unit with respect to the synchrotron radiation, on the basis of the detection performed in said detecting step; and exposing the original and the substrate with synchrotron radiation by use of the attitude adjusted exposure unit. causing a predetermined wavelength component of the synchrotron radiation to be reflected by a reflective member disposed in an exposure unit, wherein the exposure unit is capable of supporting the original and the substrate therein; detecting the reflected wavelength component; controlling the relative attitude of the exposure unit and the synchrotron radiation on the basis of the detection in said detecting step; and supplying synchrotron radiation into the exposure unit to expose the original and the substrate placed in the exposure unit. projecting a predetermined wavelength component of the synchrotron radiation onto a reflective member placed in an exposure unit which is capable of supporting the mask and the wafer therein; detecting the wavelength component as reflected by the reflective member; controlling the relative attitude of the exposure unit and the synchrotron radiation on the basis of the detection in said detecting step; and exposing the mask and the wafer, placed in the exposure unit, with synchrotron radiation, whereby the pattern of the mask is transferred to the wafer. causing a predetermined wavelength component of synchrotron radiation to be reflected by a reflective member of an exposure unit which is capable of supporting a mask and a wafer therein; detecting the wavelength component reflected by the reflective member; adjusting the attitude of the exposure unit with respect to the synchrotron radiation; and exposing the mask and the wafer in the exposure unit, with synchrotron radiation, whereby a pattern of the mask is transferred to the wafer. a radiation source for providing synchrotron radiation; an exposure unit for exposing an original and a substrate with the synchrotron radiation, said exposure unit being capable of supporting the original and the substrate therein; a reflection mirror for directing the synchrotron radiation from said radiation source to said exposure unit along a predetermined direction of projection; an alignment unit for extracting a predetermined wavelength component of light out of the synchrotron radiation and providing an alignment beam to be guided to said exposure unit; a detector for detecting the alignment beam reflected by a reflective member provided in said exposure unit; and control means for controlling the relative inclination between said exposure unit and said predetermined direction, on the basis of an output of said detector. means for providing the synchrotron radiation; an exposure unit for exposing the original and the substrate with the synchrotron radiation, said exposure unit being capable of supporting the original and the substrate therein; a reflection mirror for directing the synchrotron radiation to said exposure unit along a predetermined direction of projection; a detector for detecting an alignment beam as reflected by a reflective member provided in said exposure unit, wherein the alignment beam comprises a predetermined wavelength component of the synchrotron radiation and is directed to said exposure unit along said predetermined direction; and means for correcting any relative inclination error between said exposure unit and said predetermined direction, on the basis of an output of said detector. directing, by using a reflection mirror, the synchrotron radiation to an exposure unit along a predetermined direction of projection, wherein the exposure unit is capable of supporting the original and the substrate therein; detecting an alignment beam with a detector, wherein the alignment beam comprises a predetermined wavelength component of the synchrotron radiation as reflected by a reflective member provided in the exposure unit, and directed to the exposure unit along the predetermined direction; correcting any error in the relative attitude between the exposure unit and the predetermined direction, as defined by the reflective member, on the basis of an output of the detector; and exposing the original and the substrate supported in the exposure unit with the synchrotron radiation from the reflective member. directing, by using a reflection mirror, the synchrotron radiation to an exposure unit along a predetermined direction of projection, wherein the exposure unit is capable of supporting a mask and a wafer therein; detecting an alignment beam with a detector, wherein the alignment beam comprises a predetermined wavelength component of the synchrotron radiation as reflected by a reflective member, provided in the exposure unit, and directed to the exposure unit along the predetermined direction; correcting any error in the relative attitude between the exposure unit and the predetermined direction, as defined by the reflective member, on the basis of an output of the detector; and exposing the mask and the wafer, supported in the exposure unit, with the synchrotron radiation from the reflective member, to print a pattern of the mask on the wafer. 2. An apparatus according to claim 1, further comprising a blocking member for allowing passage of a portion of the synchrotron radiation so that the portion forms a spot on said reflective member of said exposure unit. 3. An apparatus according to claim 2, further comprising a filter for extracting a predetermined wavelength component of the synchrotron radiation, wherein the extracted wavelength component is reflected by said reflective member and is received by said detector. 4. An apparatus according to claim 3, wherein said filter is disposed upstream of said blocking member with respect to the advancement direction of the synchrotron radiation. 5. An apparatus according to claim 3, wherein said blocking member and said filter are movable out of the path of the synchrotron radiation. 6. An apparatus according to claim 3, wherein said filter extracts a visible wavelength component of the synchrotron radiation. 7. An exposure apparatus for exposing an original and a substrate with synchrotron radiation, said apparatus comprising; 8. An apparatus according to claim 7, further comprising a blocking member for allowing passage of a portion of the synchrotron radiation so that the portion forms a spot on said reflective member of said exposure unit. 9. An apparatus according to claim 8, further comprising a filter for extracting a predetermined wavelength component of the synchrotron radiation, wherein the extracted wavelength component is reflected by said reflective member and is received by said detector. 10. An apparatus according to claim 9, wherein said blocking member and said filter are movable out of the path of the synchrotron radiation. 11. An apparatus according to claim 9, wherein said filter serves to extract a visible wavelength component of the synchrotron radiation. 12. An apparatus according to claim 9, wherein said filter is disposed upstream of said blocking member with respect to the advancement direction of the synchrotron radiation. 13. An exposure method, comprising the steps of: 14. A method according to claim 13, said detecting step comprises the step of detecting a predetermined wavelength component of the synchrotron radiation for the attitude detection. 15. A method according to claim 13, said introducing step comprises the step of forming a spot on a reflective member in the exposure unit by which the synchrotron radiation is reflected. 16. An exposure method for exposing an original and a substrate with synchrotron radiation, comprising the steps of: 17. A method according to claim 16, wherein said causing step comprises the step of causing a predetermined wavelength component of the synchrotron radiation to form a spot on the reflective member, which spot has a size smaller than an illumination range to be defined by the synchrotron radiation. 18. An exposure method for transferring a pattern of a mask onto a wafer by using synchrotron radiation, said method comprising the steps of: 19. A method according to claim 18, further comprising the step of placing the mask at the position of the reflective member in place thereof, after said detecting step. 20. A semiconductor device manufacturing method using synchrotron radiation, comprising the steps of: 21. A method according to claim 20, wherein said causing step comprises the step of causing a predetermined wavelength component of synchrotron radiation to form a spot on the reflective member, which spot has a size smaller than an illumination range to be defined by the synchrotron radiation. 22. A method according to claim 20, wherein said causing step comprises the step of extracting the predetermined wavelength component of the synchrotron radiation using a filter. 23. A method according to claim 22, wherein said extracting step comprises the step of extracting a visible wavelength component of the synchrotron radiation using the filter. 24. An exposure apparatus, comprising: 25. An apparatus according to claim 24, wherein said detector is provided in said alignment unit. 26. An exposure apparatus for exposing an original and a substrate with synchrotron radiation, comprising: 27. An exposure method for exposing an original and a substrate with synchrotron radiation, said method comprising the steps of: 28. A method according to claim 27, further comprising the step of exposing the substrate with synchrotron radiation passing through a mask. 29. A method according to claim 28, further comprising the step of holding the reflective member and the mask at the same site in the exposure unit. 30. A method of manufacturing semiconductor devices by using synchrotron radiation, said method comprising the steps of:
061880736	description	EXAMPLE 1 (I) Production of Radiographic Intensifying Screen 1) Preparation of a support having light-reflecting layer containing titanium dioxide A rutile type titanium dioxide powder (500 g) having the mean grain size of 0.28 .mu.m (CR 95 [trade name], available from Ishihara Industries, Co., Ltd.) and 100 g of acrylic binder resin (Cryscoat P1018GS [trade name], available from Dainippon Ink & Chemicals, Inc.) were added into methyl ethyl ketone, and mixed to prepare a coating liquid having a viscosity of 10 PS. The coating liquid was then evenly applied by means of a doctor blade onto a polyethylene terephthalate film (thickness: 250 .mu.m) containing a titanium dioxide powder, and then dried to give a light-reflecting layer. The thickness of the dried light-reflecting layer was 40 .mu.m. The volume filling (packing) content of titanium dioxide in the support having the light-reflecting layer was 48 %, and the diffuse reflectance at a wavelength of 545 nm (which corresponds to the main peak of the luminescence emitted from terbium activated gadolinium oxysulfide Gd.sub.2 O.sub.2 S:Tb phosphor) was 95.5%. 2) Preparation of a phosphor sheet Terbium activated gadolinium oxysulfide (Gd.sub.2 O.sub.2 S:Tb, mean grain size: 3.5.mu.m, 250 g), 8 g of polyurethane binder resin (Pandex T5265M [trade name], available from Dainippon Ink & Chemicals, Inc.), 2 g of epoxy binder resin (Epikote 1001 [trade name], available from Yuka Shell Epoxy Kabushiki Kaisha) and 0.5 g of isocyanate compound (Colonate HX [trade name], available from Nippon Polyurethane Kogyo Kabushiki Kaisha) were added into methyl ethyl ketone, and mixed using a propeller mixer to prepare a coating liquid having a viscosity of 25 PS (at 25.degree. C). The coating liquid was then applied onto a temporary support (polyethylene terephthalate sheet having a surface beforehand coated with a silicon releasing agent), and dried to give a phosphor layer. The phosphor layer was then peeled off from the temporary support to prepare a phosphor sheet. 3) Fixing the phosphor sheet onto the support The above-prepared phosphor sheet was placed on the support prepared in the above 1), and then pressed by means of a calender roll at a pressure of 400 kgw/cm.sup.2 at 80.degree. C. The thickness of the resultant phosphor layer was 105 .mu.m. The volume filling content of the phosphor and the weight ratio of binder/phosphor in the phosphor layer were 68 % and 1/24, respectively. 4) Preparation of a surface protective layer Fluorocarbon resin (Lumiflon LF100 [trade name], available from Asahi Glass Co., Ltd., 10 g), 0.5 g of an alcohol modified-siloxane oligomer (X-22-2809 [trade name], available from The Shin-Etsu Chemical Co., Ltd.), 3.2 g of isocyanate (Orestar NP38-70s [trade name], available from Mitsui Toatsu Chemicals, Inc.), 0.4 g of anatase type titanium dioxide (A220 [trade name], available from Ishihara Industries Co., Ltd.; mean grain size: 0.15 .mu.m; refractive index: about 2.6) and 0.001 g of a catalyst (KS1269 [trade name], available from Kyodo Chemical Co., Ltd.) are added into a mixed solvent of methyl ethyl ketone and cyclohexanone (weight ratio: 1/1), and mixed to prepare a coating liquid. The coating liquid was then applied onto the phosphor layer by means of a doctor blade, and slowly dried. The coated layer was then heated at 120.degree. C. for 30 minutes to form a surface protective layer (thickness: 7 .mu.m). The content of titanium dioxide in the surface protective layer was 3 wt. %. (II) Calculation of the scattering length and the absorption length of the surface protective layer The coating solution of the above 4) was applied onto a transparent support (thickness: 180 .mu.m) so that the formed layer would have a thickness of 5 to 50 .mu.m. The diffuse transmittance (or diffused transmittance: %) of the formed layer was measured at a wavelength of 545 nm (corresponding to the main peak of the luminescence emitted from terbium activated gadolinium oxysulfide Gd.sub.2 O.sub.2 S:Tb phosphor), by means of an automatic recording spectrophotometer (U-3210, manufactured by HITACHI, Ltd.) equipped with an integrating sphere of 150 .phi. (150-0910). The results are set forth in Table 1. TABLE 1 thickness (.mu.m) 7 11 24 40 diffuse 70.3 62.6 48.4 40.2 transmittance (%) In accordance with the above-described formulas, the values of K and S were calculated from the data shown in Table 1. From the calculated values of K and S, the scattering length and the absorption length were determined to be 23 .mu.m (scattering length=1/S) and 10,000 .mu.m (absorption length=1/K), respectively. COMPARISON EXAMPLE 1 The procedure of Example 1 was repeated except that titanium dioxide was not added to the surface protective layer, to prepare a radiographic intensifying screen. The scattering length of the prepared screen was then determined in the same manner as described above, and found to be more than 200 .mu.m. COMPARISON EXAMPLE 2 The procedure of Example 1 was repeated except that the content of titanium dioxide powder in the surface protective layer was set to be 0.1 wt. %, to prepare a radiographic intensifying screen. The scattering length of the prepared screen was determined in the same manner as described above, and found to be 140 .mu.m. EXAMPLE 2 The procedure of Example 1 was repeated except that the content of titanium dioxide powder in the protective layer was set to be 1 wt. %, to prepare a radiographic intensifying screen. The scattering length of the prepared screen was determined, and found to be 50 .mu.m. EXAMPLE 3 The procedure of Example 1 was repeated except that the content of titanium dioxide powder in the protective layer was set to be 10 wt. %, to prepare a radiographic intensifying screen. The scattering length of the prepared screen was determined, and found to be 9 .mu.m. [Measurement of Sharpness and Sensitivity] (1) Measurement of sharpness On the surface protective layer of the sample intensifying screen, a "single emulsion layer type" radio-graphic film (X-ray film MINP 30 [trade name], available from Fuji Photo film Co., Ltd.) was overlaid so that the film might be directly contact with the protective layer. (A "single emulsion layer type" radiographic film comprises a silver halide emulsion layer provided on only one surface of its support.) The combination of the screen and the X-ray film was then exposed to X-rays through an CTF chart (made of molybdenum, thickness: 80 Am, space frequency: 0 to 10 lines/mm) in the following manner. The CTF chart was placed at a distance of 2 m from an X-ray source, and the X-ray film and the screen were placed behind the CTF chart in order. The X-ray source was composed of an X-ray generating apparatus and filters. The X-ray generating apparatus (DRX-3724HD [trade name], available from Toshiba Corporation; focal spot size: 0.6 mm.times.0.6 mm) equipped with a tungsten target and an aluminum filter (thickness: 3 mm) was activated with a three-phase pulse generator under 80 kvp, to generate X-rays. The generated X-rays were made to pass a water filter (thickness: 7 cm), which absorbed X-rays in the same amount as a human body, and then emitted from the X-ray source. After the exposure was made, the exposed film was developed in an automatic developing machine (FPM-5000 [trade name], available from Fuji Photo film Co., Ltd.) using a developer and a fixer (RD-3 and Fuji-F [trade name], respectively; available from Fuji Photo film Co., Ltd.) to obtain a sample for the measurement of sharpness. In the above exposure, the exposing conditions were adjusted so that the thick part of the resultant image would have a density of 1.8. In accordance with the method described in Japanese Patent Provisional Publication No. H9-21899, the sharpness was determined with the value at 2 lines/mm based on the obtained sample. The results are shown in Table 2. (2) Measurement of sensitivity Using the same X-ray source and the same X-ray film as described above, the combination of the screen and the X-ray film was exposed to X-rays. The distance between the X-ray source and the X-ray film was varied so that the amount of exposed X-ray might be stepwise changed (step width: logE=0.15). The exposed film was then developed in the same manner as described above to prepare a sample for the measurement of sensitivity. The density of the sample was measured with visible light to determine a characteristic curve. The sensitivity was determined with a reciprocal of the amount of exposed X-ray giving a fog density of 1.0. The sensitivity thus obtained was relatively shown so that the value of Comparison Example 1 would be 100. The results are set forth in Table 2. TABLE 2 scattering sharpness screen length (.mu.m) sensitivity (2 lines/mm) C. Ex. 1 above 200 100 0.590 C. Ex. 2 140 100 0.590 Ex. 1 23 99 0.630 Ex. 2 50 100 0.625 Ex. 3 9 97 0.615 The results shown in Table 2 indicate that each of the radiographic intensifying screens of the invention (Examples 1 to 3) gives a radiographic image having improved sharpness without lowering the sensitivity, as compared with those given by the screens of Comparison Examples 1 and 2. EXAMPLE 4 The procedure of Example 1 was repeated except that melamine resin particles (refractive index: 1.57, mean grain size: 0.6 .mu.m, content: 20 wt. %) were used in place of the titanium dioxide powder in the surface protective layer, to prepare a radiographic intensifying screen. The scattering length of the screen was determined and found to be 26 .mu.m. EXAMPLE 5 The procedure of Example 1 was repeated except that melamine resin particles (refractive index: 1.57, mean grain size: 0.6 .mu.m, content: 10 wt. %) were used in place of the titanium dioxide powder in the surface protective layer, to prepare a radiographic intensifying screen. The scattering length of the prepared screen was measured, and found to be 60 .mu.m. COMPARISON EXAMPLE 3 The procedure of Example 1 was repeated except that melamine resin particles (refractive index: 1.57, mean grain size: 3 .mu.m, content: 10 wt. %) were used in place of the titanium dioxide powder in the surface protective layer, to prepare a radiographic intensifying screen. The scattering length of the prepared screen was measured, and found to be 90 .mu.m. COMPARISON EXAMPLE 4 The procedure of Example 1 was repeated except that silicon dioxide particles (refractive index: about 1.46, mean grain size: 3 .mu.m, content: 10 wt. %) were used in place of the titanium dioxide powder n the surface protective layer, to prepare a radiographic intensifying screen. The scattering length of the prepared screen was determined, and found to be 120 .mu.m. COMPARISON EXAMPLE 5 The procedure of Example 1 was repeated except that alumina particles (refractive index: about 1.56, mean grain size: 0.8 .mu.m, content: 5 wt. %) were used in place of the titanium dioxide powder in the surface protective layer, to prepare a radiographic intensifying screen. The scattering length of the prepared screen was measured, and found to be 100 .mu.m. Measurement of Sharpness and Sensitivity (1) Measurement of sharpness The sharpness was measured in the same manner as described above. (2) Measurement of sensitivity The sensitivity was measured in the same manner as described above. The results are set forth in Table 3. TABLE 3 scattering sharpness screen length (.mu.m) sensitivity (2 lines/mm) C. Ex. 1 above 200 100 0.590 C. Ex. 3 90 100 0.600 C. Ex. 4 120 100 0.590 C. Ex. 5 100 100 0.595 Ex. 4 26 100 0.630 Ex. 5 60 100 0.620 The results shown in Table 3 indicate that each of the radiographic intensifying screens of the invention (Examples 4 and 5) gives a radiographic image having improved sharpness without lowering the sensitivity, as compared with those given by the screens of Comparison Examples 1 and 3 to 5. EXAMPLE 6 The procedure of Example 1 was repeated except that only polyethylene terephthalate was used as the binder resin to form a surface protective layer having a thickness of 6 .mu.m, to prepare a radiographic intensifying screen. The scattering length of the prepared screen was measured, and found to be 30 .mu.m. COMPARISON EXAMPLE 6 The procedure of Example 1 was repeated except that only polyethylene terephthalate was used as the binder polymer and the titanium dioxide powder was not used to form a surface protective layer having a thickness of 6 .mu.m, to prepare a radiographic intensifying screen. The scattering length of the prepared screen was measured, and found to be more than 200 .mu.m. Measurement of Sharpness, Sensitivity and Durability (1) Measurement of sharpness The sharpness was measured in the same manner as described above. (2) Measurement of sensitivity The sensitivity was measured in the same manner as described above. (3) Measurement of durability The durability of the surface protective layer was measured in the following manner. A great number of beads (diameter: 300 .mu.m) were sprinkled on a plate, and the sample intensifying screen was placed and fixed on the plate so that the support would be in contact with the beads and the surface protective layer would be pressed with the beads via the support to form a great number of convexes on the protective layer. On the surface protective layer having the convexes thus formed, a stainless steel plate (size: 4 cm.times.5 cm) and a weight of 100 g were placed and repeatedly moved so that the protective layer would be rubbed with the stainless steel plate. The rubbing had been continued until the protective layer produced crack and the phosphor layer was bared, and the times of the rubbing was counted. According to the counted rubbing times, the durability of the surface protective layer was determined. Needless to say, a large number indicates better durability. The results are set forth in Table 4. Table 4 TABLE 4 scattering sensi- sharpness dura- screen length (.mu.m) tivity (2 lines/mm) bility C. Ex. 6 above 200 98 0.595 above 10000 Ex. 6 30 97 0.635 above 10000 The results shown in Table 4 indicate that the radiographic intensifying screen of the invention gives a radiographic image having improved sharpness without lowering the sensitivity, as compared with those given by the screens of Comparison Example. Further, the results also indicate that polyethylene terephthalate binder resin gives extremely high durability to the surface protective layer. EXAMPLE 7 The procedure of Example 1 was repeated except that the thickness of the surface protective layer was set at 3 .mu.m, to prepare a radiographic intensifying screen. The scattering length of the prepared screen was measured, and found to be 23 .mu.m. EXAMPLE 8 The procedure of Example 1 was repeated except that the thickness of the surface protective layer was set at 5 .mu.m, to prepare a radiographic intensifying screen. The scattering length of the prepared screen was measured, and found to be 23 .mu.m. EXAMPLE 9 The procedure of Example 1 was repeated except that the thickness of the surface protective layer was set at 10 .mu.m, to prepare a radiographic intensifying screen. The scattering length of the prepared screen was measured, and found to be 23 .mu.m. Measurement of Sharpness and Sensitivity (1) Measurement of sharpness The sharpness was measured in the same manner as described above. (2) Measurement of sensitivity The sensitivity was measured in the same manner as described above. The results are set forth in Table 5. TABLE 5 thick- scattering sensi- sharpness screen ness (.mu.m) length (.mu.m) tivity (2 lines/mm) Ex. 7 3 23 100 0.640 Ex. 8 5 23 100 0.635 Ex. 1 7 23 99 0.630 Ex. 9 10 23 97 0.610 The results shown in Table 5 indicate that the radiographic intensifying screen of the invention gives a radiographic image having high sharpness and excellent sensitivity even if the thickness of the surface protective layer is varied. EXAMPLE 10 (1) Production of Radiographic Intensifying Screens Having Different Phosphor Layers 1) The procedure of Example 1 was repeated except that the thickness of the phosphor layer after calender treatment was set at 80 .mu.m, to prepare a radiographic intensifying screen (screen A, scattering length: 23 .mu.m). 2) The procedure of Example 1 was repeated except that 50 g of the phosphor particles having the mean grain size of 2.0 .mu.m and 200 g of those having the mean grain size of 6.2 .mu.m were used (the chemical contents of the phosphor were not changed) and that the thickness of the phosphor layer after calender treatment was set at 120 .mu.m, to prepare a radiographic intensifying screen (screen B, scattering length: 23 .mu.m). The volume filling content of the phosphor in the phosphor layer was 72%. 3) The procedure of the above 2) was repeated except that the thickness of the phosphor layer after calender treatment was set at 95 .mu.m, to prepare a radiographic intensifying screen (screen C, scattering length: 23 .mu.m). 4) The procedure of Example 1 was repeated except that a double phosphor layer consisting of a lower layer (thickness after calender treatment: 80 .mu.m) containing the phosphor particles having a mean grain size of 3.0 .mu.m and an upper layer (thickness after calender treatment: 100 .mu.m) containing those having a mean grain size of 6.2 .mu.m was formed (the chemical contents of the phosphor were not changed), to prepare a radiographic intensifying screen (screen D, scattering length: 23 .mu.m). The volume filling content of the phosphor in the phosphor layer was 70%. 5) The procedure of Example 1 was repeated except that a double phosphor layer consisting of a lower layer (thickness after calender treatment: 80 .mu.m) containing the phosphor particles having a mean grain size of 3.0 .mu.m and an upper layer (thickness after calender treatment: 240 .mu.m) containing those having a mean grain size of 6.2 Am was formed (the chemical contents of the phosphor were not changed), to prepare a radiographic intensifying screen (screen E, scattering length: 23 .mu.m). (2) Production of Silver Halide X-ray Film (Film-1) A "both-sided emulsion type" X-ray film was prepared in the same manner as described in Japanese Patent Provisional publication No. H7-219162 (sample 3 of Example 1). The subbing dye-I (described in the above publication) was applied in an amount of 45 mg per one surface. The light cross-over of the prepared film was measured by the method described in Example 1 of the above-mentioned publication, and found to be 6%. The chemical sensitization of the silver halide emulsion was adjusted so that the sensitivity and the tone might be the same as commercially available X-ray film (UR-2 [trade name], available from Fuji Photo film Co., Ltd.). (3) Evaluation of Combination of Radiographic Intensifying Screen and X-ray Film The combinations of the radiographic intensifying screens A to E and the above X-ray film (Film-1) were evaluated in the manner described in Japanese Patent Provisional publication No. H7-219162 (Example 1). In addition to that, the combinations of the screens A to E and the above commercially available X-ray film (UR-2), and those of commercially available radiographic intensifying screens (HGM2 and HGH2 [trade name], available from Kasei Optonics Co., Ltd.) and the X-ray film (UR-2) were also determined. The results are set forth in Table 6. TABLE 6 front back sensi- sharpness composition screen screen film tivity (2 lines/mm) No. 1 A B Film-1 100 0.630 No. 2 A B UR-2 100 0.580 No. 3 C D Film-1 132 0.510 No. 4 C D UR-2 132 0.470 No. 5 HGM2 HGM2 UR-2 100 0.500 No. 6 HGH2 HGH2 UR-2 130 0.410 No. 7 C E Film-1 190 0.365 No. 8 C E UR-2 190 0.315 The results shown in Table 6 indicate that the composition consisting of the screens of the invention and an X-ray film of low cross-over gives an image of improved sharpness. The results also indicate that the combination of the screens of the invention and a commercially available X-ray film gives an image having excellent balance of sensitivity and sharpness. EXAMPLE 11 1) Formation of a phosphor layer on the support having the light-reflecting layer The procedure of Example 1 was repeated except that 11 g of polyurethane binder resin was used to form a coating liquid for phosphor layer, to prepare a phosphor layer (thickness: 100 .mu.m) on the support. The volume filling content of the phosphor and the weight ratio of binder/phosphor in the phosphor layer were 66% and 1/18.5, respectively. 2) Preparation of a surface protective layer Anatase type titanium dioxide (P220 [trade name], available from Ishihara Industries Co., Ltd.) was added into melted polyethylene terephthalate (PET) resin in the amount of 3.5 wt. % (per PET resin). From thus prepared PET resin containing titanium dioxide, PET sheet (thickness: 70 .mu.m) was formed by a known extrusion method. The formed PET sheet was biaxially oriented (by 3.4 times.times.3.4 times), and then heated to prepare a thin PET film (thickness: 6.0 .mu.m) containing titanium dioxide. The diffuse transmittance of the prepared film at the wavelength of 545 nm was measured to be found 78%. Thin PET films having various thickness were also prepared in the manner described above, and then the scattering length of the PET film was measured in the same manner as described in Example 1, and found to be 25 .mu.m. The film thus prepared was overlaid and fixed on the above phosphor layer with adhesive, to provide a surface protective layer (thickness: 6.0 .mu.m). Thus, a radio-graphic intensifying screen of the invention was produced. COMPARISON EXAMPLE 7 The procedure of Example 11 was repeated except that commercially available polyethylene terephthalate film (thickness: 6 .mu.m, available from Toray Industries, Inc.) was used as the surface protective layer, to prepare a radiographic intensifying screen for comparison. The scattering length of the prepared screen was estimated to be more than 200 .mu.m. EXAMPLE 12 The procedure of Example 11 was repeated except that 15 g of polyurethane binder resin was used, to prepare a radiographic intensifying screen of the invention. The thickness of the phosphor layer, the volume filling content of the phosphor, and the binder/phosphor weight ratio in the phosphor layer were 110 .mu.m, 60%, and 1/14, respectively. COMPARISON EXAMPLE 8 The procedure of Example 12 was repeated except that commercially available polyethylene terephthalate film (thickness: 6 .mu.m, available from Toray Industries, Inc.) was used as a surface protective layer, to prepare a radiographic intensifying screen for comparison. EXAMPLE 13 The procedure of Example 11 was repeated except that 5.6 g of polyurethane binder resin and 1 g of epoxy binder resin were used, to prepare a radiographic intensifying screen of the invention. The thickness of the phosphor layer, the volume filling content of the phosphor and the weight ratio of binder/phosphor in the phosphor layer were 100 .mu.m, 70% and 1/35, respectively. COMPARISON EXAMPLE 9 The procedure of Example 13 was repeated except that commercially available polyethylene terephthalate film (thickness: 6 .mu.m, available from Toray Industries, Inc.) was used as a surface protective layer, to prepare a radiographic intensifying screen for comparison. EXAMPLE 14 The procedure of Example 11 was repeated except that 8 g of polyurethane binder resin was used, to prepare a radiographic intensifying screen of the invention. The thickness of the phosphor layer, the volume filling content of the phosphor and the binder/phosphor weight ratio in the phosphor layer were 105 .mu.m, 68% and 1/24, respectively. COMPARISON EXAMPLE 10 The procedure of Example 14 was repeated except that commercially available polyethylene terephthalate film (thickness: 6 .mu.m, available from Toray Industries, Inc.) was used as a surface protective layer, to prepare a radiographic intensifying screen for comparison. Measurement of Sharpness and Sensitivity (1) Measurement of sharpness The sharpness was measured in the same manner as described above. (2) Measurement of sensitivity The sensitivity was measured in the same manner as described above, and relatively shown so that the value of Example 14 might be 100. The results are set forth in Table 7. TABLE 7 scattering binder/phos- sensi- sharpness screen length (.mu.m) phor (wt.) tivity (2 lines/mm) Ex. 11 25 1/18.5 98 0.605 Ex. 12 25 1/14 95 0.580 Ex. 13 25 1/35 100 0.635 Ex. 14 25 1/24 100 0.635 C. Ex. 7 above 200 1/18.5 100 0.580 C. Ex. 8 above 200 1/14 99 0.570 C. Ex. 9 above 200 1/35 100 0.590 C. Ex. 10 above 200 1/24 100 0.590 The results shown in Table 7 indicate the following facts. Even if the ratio of binder/phosphor in the phosphor layer varies within the range of less than 1/12, each radiographic intensifying screen of the invention gives a radiographic image having improved sharpness without lowering the sensitivity, as compared with that given by each conventional screen having a transparent surface protective layer. Further, the screen of low binder/phosphor ratio gives good sharpness and sensitivity, and hence the ratio of binder/phosphor is preferred to be small (in other wards, the binder is preferred to be used in a small amount) in the invention. EXAMPLE 15 The procedure of Example 1 was repeated except that the thickness of the surface protective layer was set at 2 .mu.m, to prepare a radiographic intensifying screen of the invention. COMPARISON EXAMPLE 11 The procedure of Example 1 was repeated except that titanium dioxide was not used to form a surface protective layer having the thickness of 2 .mu.m, to prepare a radiographic intensifying screen for comparison. COMPARISON EXAMPLE 12 The procedure of Example 1 was repeated except that titanium dioxide was not used to form a surface protective layer having the thickness of 5 .mu.m, to prepare a radiographic intensifying screen for comparison. Measurement of Sharpness, Sensitivity, Stain Resistance and Abrasion Resistance (1) Measurement of sharpness The sharpness was measured in the same manner as described above. (2) Measurement of sensitivity The sensitivity was measured in the same manner as described above, and relatively shown so that the value of Comparison Example 1 might be 100. (3) Measurement of stain resistance 1 cc of screen cleaner (available from Fuji Photo film Co., Ltd.) was evenly applied and dried on the sample screen (size: 16 cm.times.16 cm). The thus treated sample screen and a silver halide X-ray film (UR-1 [trade name], available from Fuji Photo film Co., Ltd.) were stored at 25.degree. C., 84 %RH for 3 hours. After that, the sample screen was placed on the X-ray film so that the surface protective layer would be in contact with the film, and then pressed for fixation. The laminated screen and film were stored at 40.degree. C. for 24 hours. The screen was then peeled off from the film, and the stains caused with dyes transferred onto the protective layer from the X-ray film were observed by sight. According to the observation, the surface protective layer of each sample was classified into the following three grades: AA: not stained, PA1 BB: slightly stained, but usable, PA1 CC: stained too much to use. PA1 AA: not abraded, PA1 BB: hardly abraded and presumed to be usable even after rubbed 40,000 times, PA1 CC: slightly abraded and presumed to be usable even after rubbed 20,000 times, PA1 DD: abraded, but usable until rubbed 10000 times, PA1 EE: abraded so much that the protective layer was completely worn out and that the bared phosphor layer was stained. (4) Measurement of abrasion resistance The sample screen was rubbed 10,000 times with a UR-1 X-ray film (the rubbing film was renewed at regular intervals), and then the surface protective layer thus treated was observed by sight. According to the observation, the surface protective layer of each sample was classified into the following five grades: The results are set forth in Table 8. TABLE 8 scattering thickness sensi- sharpness length (.mu.m) binder (.mu.m) tivity (2 lines/mm) (Example 15) 23 fluoro* 2 100 0.645 stain resistance: BB abrasion resistance: DD (Example 7) 23 fluoro* 3 100 0.640 strain resistance: AA abrasion resistance: CC (Example 8) 23 fluoro* 5 100 0.635 stain resistance: AA abrasion resistance: BB (Example 1) 23 fluoro* 7 99 0.630 stain resistance: AA abrasion resistance: AA (Example 9) 23 fluoro* 10 97 0.610 stain resistance: AA abrasion resistance: AA (Comparison Example 11) above 200 fluoro* 2 100 0.635 stain resistance: BB abrasion resistance: DD (Comparison Example 12) above 200 fluoro 5 100 0.600 stain resistance: AA abrasion resistance: CC (Comparison Example 1) above 200 fluoro* 7 100 0.590 stain resistance: AA abrasion resistance: BB The results shown in Table 8 indicate that the present invention is very effective in the screen having a thick protective layer. The conventional screen having a thick protective layer gives a radiographic image of poor sharpness, while the screen of the invention having that of the same thickness gives relatively high sharpness. The results further suggests that the thick surface protective layer gives high stain resistance and high abrasion resistance. Therefore, the screen of the invention having the protective layer of enough thickness to keep sufficient resistance against stain and abrasion can give a radiographic image of high sharpness without lowering sensitivity. The results shown in Table 8 also reveal that the light-scattering particles do not lower the stain resistance but improve the abrasion resistance of the surface protective layer containing fluorocarbon resin. EXAMPLE 16 10 g of cellulose acetate (acetylation degree: about 56%) and 0.3 g of anatase type titanium dioxide (A220 [trade name], available from Ishihara Industries Co., Ltd.) were added into methyl ethyl ketone, and mixed to prepare a coating liquid for protective layer. After that, the procedure of Example 1 was repeated except that the prepared coating solution was used to prepare a surface protective layer of the thickness of 6.5 .mu.m, to prepare a radiographic intensifying screen. The content of titanium dioxide in the surface protective layer was 3 wt. %, and the scattering length of the prepared screen was 28 .mu.m. COMPARISON EXAMPLE 13 The procedure of Example 16 was repeated except that titanium dioxide was not used to form a surface protective layer, to prepare a radiographic intensifying screen for comparison. EXAMPLE 17 The procedure of Example 14 was repeated except that the thickness of the surface protective layer was set at 4 .mu.m, to prepare a radiographic intensifying screen of the invention. Measurement of Sharpness, Sensitivity, Stain Resistance and Abrasion Resistance (1) Measurement of sharpness The sharpness was measured in the same manner as described above. (2) Measurement of sensitivity The sensitivity was measured in the same manner as described above, and relatively shown so that the value of Comparison Example 10 might be 100. (3) Measurement of stain resistance The stain resistance was measured in the same manner as described above. (3) Measurement of abrasion resistance The abrasion resistance was measured in the same manner as described above. The results are set forth in Table 9. TABLE 9 scattering thickness sensi- sharpness length (.mu.m) binder (.mu.m) tivity (2 lines/mm) (Example 16) 28 cel.ac.* 6.5 100 0.630 stain resistance: BB abrasion resistance: BB (Example 17) 25 PET* 4 100 0.640 stain resistance: AA abrasion resistance: AA (Example 14) 25 PET* 6 100 0.635 stain resistance: AA abrasion resistance: AA (Comparison Example 13) above 200 cel.ac.* 6.5 100 0.585 stain resistance: BB abrasion resistance: BB (Comparison Example 10) above 200 PET* 6 100 0.590 stain resistance: AA abrasion resistance: AA (Comparison Example 1) above 200 fluoro* 7 100 0.590 stain resistance: AA abrasion resistance: BB Remark*) "cel.ac." and "PET" mean cellulose acetate and polyethylene terephthalate, respectively. The results shown in Table 9 indicate that the radiographic intensifying screen of the invention gives a radiographic image of excellent sharpness without lowering the sensitivity and the resistance against stain and abrasion even if cellulose acetate or polyethylene terephthalate is used as a binder of the surface protective layer.
051475987	abstract	A nuclear reactor core has a first group of fuel rods containing fissionable material and no burnable absorber, and a second group of fuel rods containing fissionable material and two burnable absorber materials. The groups of fuel rods are arranged in the core for controlling power peaking and moderator temperature coefficient. The number of fuel rods in the first group are greater than the number in the second group. The two burnable absorber materials can be provided as separate coatings or a mixture. One burnable absorber material is an erbium-bearing material such as erbium oxide and the other is a boron-bearing material such as zirconium diboride. Alternatively, the erbium-bearing material can be interspersed or mixed with the fissionable material.
	abstract	An x-ray apparatus (1), has an electron beam source (2), a target (4), onto which the electron beam (3) is directed to form a focal spot (5; 5a, 5b) on the target (4), x-ray optics (6) for collecting x-rays emitted from the focal spot (5; 5a, 5b) to form an x-ray beam (8) and a sample position (9) at which the x-ray beam (8) is directed. The x-ray apparatus (1) further includes an electrostatic or electromagnetic electron beam deflection device (10) suitable for moving the focal spot (5; 5a, 5b) on the target (4). The extension of the focal spot (5; 5a, 5b) in any direction (x, y, z) is at least a factor of 1.5 smaller than the extension of the target (4). An x-ray apparatus is thereby provided with simplified alignment of the x-ray optics with respect to a microfocus x-ray source.
	description	A preferred embodiment of a nuclear reactor in accordance with the present invention will be described below. FIG. 1 and FIG. 2 are vertical cross-sectional views showing reactor containment vessels to which the first embodiment is applied. A pressure vessel 1 containing a reactor core 3 constructed of nuclear fuel, a shroud 2 and control rods 7 is contained in a containment vessel 11. The containment vessel 11 is composed of a drywell 12 to install the pressure vessel 1 therein; a pressure suppression pool 14 for suppressing pressure of the containment vessel 11 by condensing steam through vent pipes 18 in an event of a reactor accident such as occurrence of rupture in a main steam pipe 5; and a wetwell 16 communicating with an upper plenum 15 of the pressure suppression pool 14 through a communicating pipe 19. By arranging a pipe having an automatic depressurizing valve 21 between the inside of the pressure vessel 1 and a quencher 22, steam inside the pressure vessel 1 can be discharged to the pressure suppression pool 14 through the quencher 22. The containment vessel 11 made of reinforced concrete except a region installing the pressure vessel 1 is vertically partitioned into three compartments, the pressure suppression pool 14 having cooling water being formed in the upper compartment, the drywell 12 for arranging components such as the main steam pipe 5 and so on being formed in the middle compartment, the wetwell 16 of pressure suppression space being formed in the lower compartment, the drywall 12 communicating with the pressure suppression pool 14 through the plurality of vent pipes 18. Further, a plurality of gravitationally flow-down water injection pipes 24 having an isolation valve 23 are arranged between the pressure vessel 1 and the pressure suppression pool 14. A cooling vessel 31 filled with a coolant is arranged inside the containment vessel 11 in a position at a level higher than that of the reactor core 3 inside the pressure vessel 1, and a heat exchanger 34 is arranged in a position at a level lower than the liquid surface level of the cooling vessel 31, and the heat exchanger 34 communicates with an inside portion of the pressure vessel 1 at a level lower than the water surface level in the pressure vessel during normal operation through a pressure vessel water injection pipe 35 and an inflow pipe 33. A heat dissipater 37 is arranged in a position at a level higher than that of the cooling vessel 31 and inside a ventilation duct 39 outside the reactor building, and the heat dissipater 37 communicates with the upper portion of the cooling vessel 31 through a gas inflow pipe 36, and the heat dissipater 37 communicates with the inside of the cooling vessel 37 through a liquid returning pipe 38. Further, a heat-pipe type containment vessel cooling system is constructed by arranging a condensing type heat exchanger 41 filled with a heat medium in an upper plenum 15 of the pressure suppression pool 14 and a heat dissipater 43 at a level higher than that of the condensing type heat exchanger 41 in the ventilation duct 39 outside the containment vessel 11, and by making the condensing type heat exchanger 41 communicate with the heat dissipater 42 through a liquid returning pipe 42 and a gas inflow pipe 44. A system composed of pressure vessel bottom water flooding pipes 51, an isolation valve 52 and a fuse valve 53 is a system for keeping a reactor-core melted substance inside the pressure vessel 1 in an event of occurrence of such a severe accident that the reactor core is melted down onto the bottom head of the pressure vessel 1 though such a severe accident hardly occurs, and the system cools the outer surface of the bottom head of the pressure vessel 1 by injecting cooling water of the pressure suppression pool 14 into the lower portion of the drywell 13 through the pressure vessel bottom water flooding pipe 51. In the inside of the reactor containment vessel 11 described above, a heat exchanger 72 is arranged in the pressure suppression pool 14. A main steam pipe 5 is branched between a main steam isolation valve 25 and a heat exchanger 4, and a steam pipe 68 with an isolation valve 66 for making the main steam pipe 5 communicate with the heat exchanger 72 is arranged. Further, a feedwater pipe 6 is branched between an isolation valve 65 and a heat exchanger 4, and a cooling water returning pipe 69 with an isolation valve 67 for making the feedwater pipe 6 communicate with the heat exchanger 72 is arranged. In a case where the main condenser can not be used during reactor shutdown, or in a case where an accident of bringing the inside of the pressure vessel into an overheated state occurs, the main stream isolation valve 25 and the isolation valve 65 are closed and the isolation valve 66 and the isolation valve 67 are opened. Decay heat in the reactor primary system is removed by boiling of cooling water in the secondary system of the heat exchanger 4, and steam of the secondary system flows from the steam pipe 68 into the heat exchanger 72 to be cooled and condensed by water inside the pressure suppression pool 14. On the other hand, the condensed water of the heat exchanger 72 flows into the cooling water returning pipe 69 to be supplied to the heat exchanger 4. The decay heat in the reactor primary system is transferred to the water in the pressure suppression pool 14 through the heat exchanger 72 to increase temperature of the water in the pressure suppression pool 14. Steam is generated when the water temperature of the pressure suppression pool 14 exceeds the saturation temperature, and the steam in the upper plenum 15 of the pressure suppression pool 14 is condensed by the condensing type heat exchanger 41. Therefore, the pressure suppression pool 14 is cooled. The decay heat in the reactor primary system is discharged outside the containment vessel by the heat dissipater 43 which communicates with the condensing type heat exchanger 41 through the liquid returning pipe 42 and the gas inflow pipe 44. Thereby, the decay heat in the reactor primary system can be removed without letting the cooling water of the reactor primary system into the containment vessel. Since the present embodiment can remove the decay heat in the reactor primary system without using any active components such as a pump or the like, there is no need to provide the isolation cooling pool and the shielding structure which have been arranged outside the containment vessel in a conventional boiling water reactor. Therefore, the economic feasibility, the reliability and the safety of the nuclear reactor can be improved. A second embodiment in accordance with the present invention will be described below, referring to FIG. 3. The second embodiment is that in the nuclear reactor shown in the first embodiment, the main steam pipe 5 is branched between the main steam isolation valve 25 and the heat exchanger 4, and a steam pipe 68 with an isolation valve 66 for making the main steam pipe 5 communicate with the pressure suppression pool 14 is arranged. A quencher 71 for moderating pressure fluctuation during steam condensing is arranged in the outlet of the steam pipe 68. Further, a feedwater pipe 6 is branched between an isolation valve 65 and a heat exchanger 4, and a cooling water returning pipe 69 with an isolation valve 67 for making the feedwater pipe 6 communicate with the pressure suppression pool 14 is arranged. Although the decay heat in the reactor primary system needs to be removed during reactor shutdown, there is a possibility that the main condenser can not be used at that time though the possibility is very low. The conventional reactor has a cooling system using an active component for taking such a case into consideration. In the present embodiment, the system for removing the decay heat in the primary system using the heat exchanger in accordance with the present invention will be described below. In a case where the main condenser can not be used during reactor shutdown, the main stream isolation valve 25 and the isolation valve 65 are closed and the isolation valve 66 and the isolation valve 67 are opened. Decay heat in the reactor primary system is removed by boiling of cooling water in the secondary system of the heat exchanger 4, and steam of the secondary system flows from the steam pipe 68 into the heat exchanger 72 to be cooled and condensed by water inside the pressure suppression pool 14. On the other hand, the condensed water of the heat exchanger 72 flows into the cooling water returning pipe 69 to be supplied to the heat exchanger 4. Thereby, the decay heat in the reactor primary system can be removed without letting the cooling water of the reactor primary system into the containment vessel. Since the present embodiment can remove the decay heat in the reactor primary system without using any active components such as a pump or the like, the economic feasibility, the reliability and the safety of the nuclear reactor can be improved. The other constructions and functions are the same as those of the first embodiment. An example of the structure inside a reactor pressure vessel suitable for being employed in each of the embodiments in accordance with the present invention will be described below as a third embodiment. FIG. 4 is a vertical cross-sectional view showing the inside of the pressure vessel to which the third embodiment is applied, and FIG. 5 is a horizontal cross-sectional view showing the inside of the pressure vessel, and FIG. 6 is a vertical cross-sectional view showing a heat exchanger tube, and FIG. 7 is a diagram showing the heat balance of the primary system cooling water and the secondary system cooling water in a region between the heat exchanger feed-water header and the steam header. The structure of the reactor pressure vessel employed in the first embodiment and the second embodiment is as follows. An annular baffle plate 10 having a flow cross-sectional area smaller than a flow area of the shroud 2 is arranged above the shroud 2 inside the pressure vessel 1 of the nuclear reactor, and the heat exchanger 4 is arranged outside the baffle plate 10. The heat exchanger 4 is placed in a position at a level higher than that of a water level inside the pressure vessel during the normal operation, and accordingly steam of the primary cooling water is condensed in the heat exchanger 4 to transfer the heat to the secondary cooling water. Further, a gap is formed between the upper portion of the shroud 2 and the baffle plate 10 to form a flow passage 26, and consequently part of the primary cooling water heated by the reactor core 3 and flowing upward inside the shroud 2 circulates by flowing down from the flow passage 26 to a downcomer 80. The secondary steam generated in the heat exchanger 4 is transferred from the main steam pipe 5 to the outside of the containment vessel 11 through a steam header 28 and the main steam isolation valve 25 to be used for driving a turbine for electric power generation or used for purpose of heat supply. Feed water from the outside of the containment vessel 11 is supplied from the feed-water header 27 to the heat exchanger 4 through a feed water pipe 6 and the isolation valve 65. By extracting and supplying the secondary cooling water through the upper head 79 of the pressure vessel, it is possible to prevent an event in relating to loss of primary cooling water from occurring because there is no large diameter pipe in a position at a level lower than the water surface level of the primary cooling water in the pressure vessel 1. The horizontal cross-sectional view of FIG. 5 shows the cross section on the plane of the line Axe2x80x94Axe2x80x2 of FIG. 4. The heat exchanger 4 is composed of four heat exchanger sectional units. The number of the heat exchanger sectional units is equal to number of the main stream pipe 5 lines, and each of the heat exchanger sectional units of the heat exchanger 4 is allocated to and connected to each of the main stream pipe 4 lines to individually form a system. By arranging the plurality of heat exchanger 4 sectional units as described above, even in an event of occurrence of a rupture in one of the secondary pipe lines such as the main steam pipe 5 or the feed water pipe 6, cooling of the primary cooling water in the pressure vessel 1 can be continued using the other systems in which no rupture occurs in the line. The tube 29 of the heat exchanger 4 is shown in the vertical cross-sectional view of FIG. 6. In the primary side of the heat exchanger 4, the primary system steam flowing from the upper portion is condensed on the upper portion of the heat exchanger tube 29, and the primary system condensed water 30 flows downward to the lower portion of the heat exchanger tube in a form of liquid film. In the secondary side of the heat exchanger 4, the single phase secondary system feed water 40 flows in from the lower portion, and is heated and boiled by the primary system flowing-down liquid film, and is further heated by condensing heat transfer of the primary system to be turned into secondary system steam. Since the heat transfer is performed by condensing and liquid film heat transfer in the primary system and by boiling heat transfer in the secondary system, high efficiency heat exchange can be performed. By forming the outer diameter of the baffle plate 10 smaller than the diameter of the shroud 2, the wide installation room for the heat exchanger 4 is secured in the annular space between the baffle plate 10 and the pressure vessel 1, as shown in FIG. 5, and consequently the heat exchanging heat transfer area of the heat exchanger 4 can be secured wider. FIG. 7 shows the concept of heat balance in the present embodiment of the nuclear reactor. In the present embodiment, the flow pattern and the cooling water temperature of the primary system and the secondary system were calculated using dimensions of a typical boiling water reactor and under conditions of primary system pressure of 12.3 MPa, secondary system pressure of 7.1 MPa, thermal output of the reactor core of 434 MWt, height of the heat exchanger 4 of 4 m, and heat transfer area of 2500 m2. When temperature of the primary system steam at the inlet of the heat exchanger 4 is set to 598 K, temperature of the primary system cooling water at the outlet becomes 593 K by heat exchange. On the other hand, when temperature of the secondary system cooling water at the inlet is set to 489 K equivalent to the feed water temperature of the existing boiling water reactor, secondary system steam having temperature of 559 K can be obtained. According to the present embodiment, since the heat transfer is performed by condensing and liquid film heat transfer in the primary system and by boiling heat transfer in the secondary system, there is an effect in that economic feature of the nuclear reactor can be improved. Examples of the structure inside a reactor pressure vessel applicable to each of the first embodiment and the second embodiment in accordance with the present invention will be described below as a fourth embodiment and a fifth embodiment. FIG. 8 is a vertical cross-sectional view showing the reactor pressure vessel 1 to which the fourth embodiment is applied, and FIG. 9 is a vertical cross-sectional view showing the reactor pressure vessel 1 to which the fifth embodiment is applied, and FIG. 10 is a horizontal cross-sectional view showing the heat exchanger 4. The present embodiment is that in the pressure vessel 1 shown in the third embodiment, a superheater 61 and a heat exchanger 4, which are heated by the primary cooling water circulating through the reactor core 3, are arranged inside the baffle plate 10. Referring to FIG. 8 and FIG. 9, the pressure vessel 1 is of a type of inserting control rods 7 from the upper portion, and control rod drive mechanisms 8 are attached to the top head 79 of the vessel. By making both of or either of the superheater 61 and the heat exchanger 4 supported by the top head 79 of the pressure vessel together with control rod drive shafts of the control rod drive mechanisms 79, when both of or either of the superheater 61 and the heat exchanger 4 is taken off at maintenance of the pressure vessel 1, both of or either of the superheater 61 and the heat exchanger 4 can be easily taken off from the pressure vessel 1 by pulling up both of or either of the superheater 61 and the heat exchanger 4 together with the top head 79 of the pressure vessel 1. In the pressure vessel 1 shown in FIG. 8, the superheater 61 is arranged inside the baffle plate 10. The secondary cooling water flowing from the feed water header 27 into the heat exchanger 4 is heated to be changed to steam and reaches a steam header 28. Then, the steam is further superheated by the primary cooling water circulating through the reactor core 3 while the steam is flowing down in the superheater 61, and becomes high-quality steam having a less moisture content to flow out from a steam header 62 to the main steam pipe 5. In the pressure vessel 1 shown in FIG. 9, the heat exchanger 70 is arranged inside the baffle plate 10. The secondary cooling water flowing from the feed water header 27 into the heat exchanger 4 is heated in the heat exchanger 4 and the heat exchanger 70 to be changed to steam and reaches the steam header 28. Then, the steam flows out to the main steam pipe 5. In the present embodiment, there is an effect that the safety of the nuclear reactor is improved because it is possible to prevent an event in relating to loss of primary cooling water from occurring. Further, there is an effect that the safety of the nuclear reactor can be improved because cooling of the primary cooling water can be continued in an event of occurrence of a rupture in the secondary system pipe. Further, there is an effect that the maintainability of the nuclear reactor can be improved because when both of or either of the superheater 61 and the heat exchanger 4 is taken off at maintenance of the pressure vessel, both of or either of the superheater 61 and the heat exchanger 4 can be taken off together with the top head 79 of the pressure vessel 1. Furthermore, in the embodiment shown in FIG. 8, there is an effect that the economic feasibility of the nuclear reactor can be improved because moisture content in the secondary steam can be reduced to improve the thermal efficiency. In the embodiment shown in FIG. 9, there is an effect that the economic feasibility of the nuclear reactor can be improved because the total heat transfer area of the heat exchanger can be increased to increase the output power of the nuclear reactor. An example of the structure inside a reactor pressure vessel 1 applicable to each of the first embodiment and the second embodiment in accordance with the present invention will be described below as a sixth embodiment. FIG. 11 is a vertical cross-sectional view showing the reactor pressure vessel to which the sixth embodiment is applied. That is, in the pressure vessels shown by the third embodiment and the fourth embodiment, a preheater 63 is arranged in a position at a level lower than the heat exchanger 4 and lower than the water level of the cooling water of the downcomer 80. The secondary cooling water flowing from the feed water header 27 into the preheater 63 is heated by the primary cooling water inside the downcomer 80, and then the secondary cooling water in an easily boiling state of a small subcooling degree flows into the heat exchanger 4. After that, the secondary cooling water is heated by the heat exchanger 4 to change into steam and reaches the steam header 28, and then flows out to the main steam pipe 5. Since the boiling heat transfer region in the heat exchanger 4 is increased, heat transfer of the secondary cooling water is improved to decrease the moisture content in the steam. Further, by separating the feed water header of the preheater 63 from the feed water header of the heat exchanger 4, and by controlling feed water flow rates of the both systems, an amount of transferred heat of the primary cooling water in the downcomer 80 can be controlled. Thereby, since the subcooling degree of the primary cooling water at the reactor core inlet can be controlled, the operability of the nuclear reactor can be improved. In the present embodiment, since the heat transfer performance of the heat exchanger can be improved to decrease the moisture content in the secondary steam, there is an effect in that the thermal efficiency can be improved and the economic feasibility of the nuclear reactor can be improved. Further, since the subcooling degree of the primary cooling water at the reactor core inlet can be controlled, there is an effect in that the operability of the nuclear reactor can be improved. In order to employ the third embodiment, the fourth embodiment or the fifth embodiment of the pressure vessel to the first embodiment or the second embodiment, the main steam pipe 5 after projecting upward from the pressure vessel 1 is branched to an upward branched main steam pipe 5 and a downward branched main steam pipe 5. The upward branched main steam pipe 5 is connected to the isolation valve 66 and then connected to the heat exchanger 72 through a steam pipe 68. On the other hand, the downward branched main steam pipe 5 is connected to the main steam isolation valve 25. Further, the feed water pipe 6 after projecting upward from the pressure vessel 1 is branched to an upward branched feed water pipe 6 and a downward branched feed water pipe 6. The upward branched feed water pipe 6 is connected to the isolation valve 67 and then connected to the heat exchanger 72 through the cooling water returning pipe 69. On the other hand, the downward branched feed water pipe 6 is connected to the isolation valve 65. An example of the structure inside a reactor pressure vessel 1 applicable to each of the first embodiment and the second embodiment in accordance with the present invention will be described below as a seventh embodiment. FIG. 12 is a vertical cross-sectional view showing the reactor pressure vessel to which the seventh embodiment is applied. FIG. 12 shows an example in which the present embodiment is applied to the third embodiment of the reactor pressure vessel. As shown in FIG. 12, in the reactor pressure vessel 1 shown in the third embodiment, a baffle plate 73 and a baffle plate 74 are placed inside the baffle plate 10 so as to intersecting at right angle with the flow direction of the two-phase cooling water flow flowing from the inside of the shroud 2 into the baffle plate 10. The baffle plate 73 and the baffle plate 74 individually have a plurality of flow passage holes 75, 76, respectively, and the flow passage holes 75 in the baffle plate 73 and the flow passage holes 76 in the baffle plates 74 are formed at arrangement positions so as to not vertically overlapped with one another. The primary cooling water in a two-phase flow flowing up inside the baffle plate 10 collides against the baffle plate 74 to change the flow direction to the horizontal direction, and then part of the primary cooling water flows into the gap between the baffle plate 73 and the baffle plate 74 through the flow passage holes 76. Since the positions of the flow passage holes 76 and the flow passage holes 75 are different from one another, the cooling water flowing into the gap through the flow passage holes 76 collides against the baffle plate 73. After that, the cooling water changes the flow direction to the horizontal direction, and then passes through the flow passage holes 75 to the plenum above the baffle plate 73. Steam separation of the two-phase primary cooling water is accelerated by the collision and the flow direction change from flowing inside the baffle plate 10 to passing through the baffle plate 73 to the upper portion. Thereby, the moisture content in the primary cooling water reaching the heat exchanger 4 from the baffle plate 73 is decreased to improve the heat transfer performance of the heat exchanger. In order to apply the seventh embodiment of the reactor pressure vessel 1 to the first embodiment or the second embodiment, the main steam pipe 5 in the sixth embodiment is branched into two pipes at a position outside the reactor pressure vessel 1, and one of the branched pipes is connected to the isolation valve 66, and the other of the branched pipes is connected to the main steam isolation valve 25. On the other hand, the feed water pipe 6 is branched into two pipes at a position outside the reactor pressure vessel 1, and one of the branched pipes is connected to the isolation valve 67, and the other of the branched pipes is connected to the main steam isolation valve 65. Similarly, FIG. 13 shows an eighth embodiment, and FIG. 14 shows the horizontal cross-sectional view. The embodiment eighth is an example in which the baffle plate 73 and the baffle plate 74 of the seventh embodiment are added to the fourth embodiment so as to arrange them below the superheater 61 as in the sixth embodiment. Since the structure inside the reactor pressure vessel of FIG. 13 is that in the structure shown in FIG. 11, the baffle plate 73 and the baffle plate 74 are arranged inside the baffle plate 10 so as to intersect at right angle with the flow direction of the two-phase cooling water flowing from the inside of the shroud 2 into the inside of the baffle plate 10, the eighth embodiment has the same operation and the same effect to the siperheater 61 as those of the seventh embodiment shown in FIG. 12. In the present embodiment, there is an effect in that the economic feasibility of the nuclear reactor can be improved because the heat transfer performance of the heat exchange is improved and accordingly the moisture content in the secondary steam is decreased to improve the thermal efficiency. In order to apply the eighth embodiment of the reactor pressure vessel 1 to the first embodiment or the second embodiment, the main steam pipe 5 and the feed water pipe 6 are individually connect to the isolation valves, similarly to the case of the fourth embodiment. A matter relating to assembling of the shroud 2 and the baffle plate 10 to the reactor pressure vessel 1 in each of the embodiments described above will be described below. Although the description will be made by taking the fifth embodiment in accordance with the present invention as an example, the description is applicable to each of the other embodiments. The shroud 2 surrounding the reactor core 3 is divided into a plurality of parts, for example, into a lower shroud 91 and an upper shroud 92. The lower shroud 91 and the upper shroud 92 are vertically joined together with bolts 93. The upper shroud 92 and the baffle plate 10 are vertically joined together with bolts 94. Therefore, the baffle plate 10, the upper shroud and the lower shroud 91 can be split into individual parts by unfastening the bolts 93, 94. By the structure described above, the reactor core 3 and the control rods 7 can be taken out without interfering with the other parts at disassembling the core internals during scheduled inspection of the nuclear reactor by removing the baffle plate 10, the upper shroud 92 and the lower shroud 91 in this order from the top. Since disassembling work time during the scheduled inspection of the nuclear reactor can be shortened by employing the above-described assembling structure of the baffle plate 10, the upper shroud 92 and the lower shroud 91, there is an effect in that a period of the scheduled inspection can be shortened and accordingly the economic feature of the nuclear reactor can be improved. The reactor pressure vessel 1 according to any one of the third embodiment to the eighth embodiment using the shroud 2 and baffle plate 10 having the above-described assembling structure is installed inside the containment vessel 11 having the decay heat removal system described in the first embodiment or the second embodiment. Here, it is assumed that an accident of the main steam pipe 5 rupture or the feed water pipe 6 rupture occurs inside the containment vessel 11 which contains any one of the reactor pressure vessels described above. Initially, both of the isolation valves 26, 65 are closed to isolate influence of the accident from the outside of the containment vessel though during normal operation both of the isolation valves 26, 65 have been opened so as to supply steam to the turbine for electric power generation through the main steam pipe 5 as the driving steam and so as to return condensate water of the used steam to the heat exchanger 4 through the feed water pipe 6. At the same time, both of the isolation valves 66, 67 are opened to dissipate decay heat from the inside of the pressure vessel 1 into the pressure suppression pool 14, as described in the descriptions of the first embodiment and the second embodiment. It is difficult to transfer the heat inside the pressure vessel 1 from the heat exchanger 4 connected to the broken main steam pipe 5 or the broken feed water pipe 6 to the side of the quencher 22 or the heat exchanger 72 illustrated in FIG. 3 which is also connected to the broken main steam pipe 5 or the broken feed water pipe 6, respectively. However, since between the pressure vessel 1 and the pressure suppression pool 14, there are provided the plurality of decay heat removal systems from the main steam pipe 5 or the feed water pipe 6, and the heat exchanger 4 to the heat exchanger 72 or the quencher 22 in FIG. 3, removing of decay heat can be performed using the unbroken decay heat removal systems. Therein, in the case of the second embodiment, both of the isolation valves 66, 67 communicating with the broken main steam pipe 5 are closed to prevent the water in the pressure suppression pool 14 from leaking through the main steam pipe 5 or the feed water pipe 6. When an accident of the main steam pipe 5 rupture or the feed water pipe 6 rupture occurs, an incondensable gas in the drywell 12 initially flows into the pressure suppression pool 14 through the vent pipes 18, and then flows into the wetwell 16 together with an incondensable gas existing in the upper plenum of the pressure suppression pool 14 through the communicating pipe 19. After that, steam flowing out through the portion of pipe rupture flows into the pressure suppression pool 14 to be condensed by the pool water. The upper plenum of the pressure suppression pool 14 is filled with the incondensable gas remaining in the upper portion of the pool and steam having a steam vapor pressure corresponding to the saturation pressure of the pool water. As inflow of the steam is further continued, the incondensable gas having a lighter specific weight flows into the wetwell 16 through the communicating pipe 19. At that time, the pressure in the pressure suppression pool, that is, the pressure in the containment vessel becomes a value of the sum of the pressure of incondensable gas pressure and the partial pressure of the steam in a conventional nuclear reactor. However, in the nuclear reactor in accordance with the present invention, because the incondensable gas is separated to the wetwell and the steam is separated to the pressure suppression pool 14, the pressure in the containment vessel becomes a higher pressure between the both partial pressures. Therefore, the pressure in the containment vessel 11 becomes a value of the sum of the incondensable gas partial pressure and the steam partial pressure, and accordingly the pressure in the containment vessel 11 can be suppressed to be increased. When the water temperature of the pressure suppression pool 14 is low, the incondensable gas remains in the upper plenum of the pressure suppression pool 14 because the partial pressure of the incondensable gas is high, and the pressure in the containment vessel becomes equal to the pressure of the incondensable gas. On the other hand, when the water temperature of the pressure suppression pool 14 is high, the pressure in the containment vessel becomes equal to the partial pressure of the steam and part of the steam flows into the wetwell 16 because the partial pressure of the steam is high. Calculating a case where the volume of the wetwell 16 is equal to the sum of the volumes of the drywell 12 and the lower drywell 13, the partial pressure of the incondensable gas becomes approximately 2 atmospheres and the partial pressure of the steam becomes approximately 5 atmospheres at pool water temperature of 160 degrees. Threfore, in the conventional technology, the pressure in the containment vessel becomes approximately 7 atmospheres. On the other hand, in the present invention, the pressure in the containment vessel is suppressed to approximately 5 atmospheres and accordingly the pressure can be reduced by 2 atmospheres. As described above, the safety of the nuclear reactor can be improved because the pressure increase at occurrence of the accident can be suppressed, and the economic feature in relation to manufacturing of the nuclear reactor can be improved because the design pressure of the containment vessel can be reduced and the strength of the structural material can be optimized. Since discharging of the steam through the broken port in an event of accident is equivalent to discharging of the cooling water outside the pressure vessel 1, the water level on the pressure vessel 1 is decreased. Thereby, since the lower ends of the pressure vessel water injection pipe 35 and the heat exchanger inflow pipe 33 become higher than the water level in the pressure vessel 1, the cooling water filled in the pressure vessel water injection pipe 35 and the heat exchanger inflow pipe 33 and the heat exchanger 34 flow down into the pressure vessel 1, and the insides of the pressure vessel water injection pipe 35 and the heat exchanger inflow pipe 33 and the heat exchanger 34 are filled with steam instead. In the heat exchanger 34, the steam is heat-removed and condensed by the cooling water in the cooling vessel 31, and the condensed water flows down through the pressure vessel water injection pipe 35 to be injected into the pressure vessel 1. The steam in the pressure vessel 1 is newly sucked through the heat exchanger inflow pipe 33 by flowing-down of the condensed water. Thus, the condensation in the heat exchanger 34 and the injection of the condensed water into the pressure vessel 1 are continued. On the other hand, the cooling water in the cooling vessel 31 boils because heat is transferred to the cooling water in the cooling vessel 31 by condensation in the heat exchanger 34. The steam generated by the boiling flows into the gas inflow pipe 36 to be condensed in the heat dissipater 37 cooled by atmospheric air inside the ventilation duct 39 outside the reactor building. The condensed water in the heat dissipater 37 is circulated to the cooling vessel 31 through the liquid returning pipe 38. Thereby, the heat generated in the reactor core 3 transferred to atmospheric air in natural convection inside the ventilation duct 39 through the heat exchanger 34, the cooling vessel 31 and the heat dissipater 37. The heat removal described above can be attained using only the piping and the heat exchangers without using any active components such as a pump, a valve and the like, and not limited by an amount of cooling water, and the heat dissipation by atmospheric air can be permanently continued. Therefore, long-term cooling can be performed until the accident event is completely settled. In addition, since the heat removal system is of a double isolation structure that the steam flowing from the pressure vessel 1 is isolated by the heat exchanger 37 and the cooling water in the cooling vessel 31 is isolated by the heat dissipater 37, it is possible to doubly prevent the primary cooling water having radioactivity from flowing out to the outside of the containment vessel 11. Description will be made below on operation of the structure that the condensing type heat exchanger 41 filled with a heat medium is arranged in the upper plenum 15 of the pressure suppression pool 14, and the heat dissipater 43 is arranged outside the containment vessel 11, and the condensing type heat exchanger 41 and the heat dissipater 43 are made to communicate with each other through the liquid returning pipe 42 and the gas inflow pipe 44. The steam flowing out to the drywell 12 at an nuclear reactor accident such as rupture of the main steam pipe 5 or the like is flows from the vent pipes 18 into the pressure suppression pool 14 to be condensed. The latent heat of the steam is transferred to the cooling water of the pressure suppression pool 14 by the condensation to increase temperature of the cooling water and pressure in the upper plenum 15. In order to suppress increase of the pressure in the upper plenum 15, it is necessary to cool the inside of the upper plenum 15 over a long term. In the heat pipe type containment vessel cooling system in the first embodiment, the steam in the upper plenum 15 is condensed and heat removed using the condensing type heat exchanger 41. The heat medium in the condensing type heat exchanger 41 is heated and vaporized, and flows from the inside of the gas inflow pipe 44 to the heat dissipater 43, and then cooled and condensed by the atmospheric air in natural convection inside the ventilation duct 39 outside the reactor building. The condensed water is circulated to the condensing type heat exchanger 41 through the liquid returning pipe 42. Thereby, the heat generated in the reactor core 3 transferred to atmospheric air inside the ventilation duct 39 outside the reactor building through the drywell 12, the vent pipes 18, the upper plenum 15, the condensing type heat exchanger 41, the heat dissipater 43, and thus the containment vessel 11 is cooled. The heat removal described above can be attained using only the piping and the heat exchangers without using any active components such as a pump, a valve and the like, and not limited by an amount of cooling water. Further, long-term cooling can be performed until the accident event is completely settled because the heat dissipation can be permanently continued. Further, since the employed heat transfer modes is vaporization and condensation of the cooling medium and not natural convection heat transfer, the heat transfer is better and accordingly a high heat dissipation efficiency can be obtained. Furthermore, in an event of a single rupture of breaking of a pipe in the condensing type heat exchanger 41 in the upper plenum 15, the coolant inside the containment vessel 11 does not flow out to the outside of the containment vessel. Therefore, the safety and the reliability of the nuclear reactor can be improved. In addition, since the incondensable gas flows into the wetwell 16 to reduce the concentration of the incondensable gas in the upper plenum 15, heat transfer of the condensing type heat exchanger 41 installed in the upper plenum 15 becomes better and accordingly size of the condensing type heat exchanger 41 can be made small. Description will be made below on operation of the structure that the plurality of gravitationally flow-down water injection pipes 24 having the isolation valve 23 are arranged between the pressure vessel 1 and the pressure suppression pool 14, and the plurality of pressure vessel bottom water flooding pipes 51 are between the pressure suppression pool 14 and the drywell 12, and the upper ends of the gravitationally flow-down water injection pipes 24 and the upper ends of the pressure vessel bottom water flooding pipes 51 are arranged at a level higher than a level of the outlets of the vent pipes 18 in the pressure suppression pool 14. At a reactor accident such as occurrence of a rupture in the main steam pipe 5, the cooling water in the pressure suppression pool 14 can be injected into the pressure vessel 1 by opening the isolation valve 23 when cooling of the inside of the pressure vessel 1 is progressed and the pressure in the pressure vessel 1 is decreased. Further, when the pressure in the pressure vessel is still high, only the automatic depressurizing valve 21 communicating with the sound main steam pipe 5 without occurrence of pipe rupture is opened, and consequently the steam inside the pressure vessel 1 is injected from the quencher 22 of FIG. 1 into the pressure suppression pool 14 to be condensed. By doing so, when the pressure in the pressure vessel 1 is reduced by releasing the pressure in the pressure vessel outside the pressure vessel, by opening the isolation valve 23 the water stored in the pressure suppression pool 14 is injected into the pressure vessel 1 through the gravitationally flow-down water injection pipes 24 to cool the inside of the pressure vessel. Further, if the isolation valve 23 of the gravitationally flow-down water injection pipe 24 could not be opened and consequently the gravitationally flow-down water injection system could not be operated, the isolation valve 52 of the pressure vessel bottom water flooding pipe 51 would be opened to inject the cooling water in the pressure suppression pool 14 into the lower drywell 13. As the water injection is started, the gas in the lower drywell 13 is blown off into the drywell 12 to fill the lower drywell 13 with the injected cooling water. Thus, the outside of the bottom head of the pressure vessel 1 is emerged under the cooling water to cool the inside of the pressure vessel 1 through heat conduction in the wall of the lower hear. If the water level of the cooling water in the pressure vessel would be further lowered to expose the reactor core to vapor, or if all the valves could not be operated and the cooling water could not be injected into the pressure vessel 1 and the lower drywell 13 to expose the reactor core to steam, a severe accident that the core would be melted down onto the bottom head of the pressure vessel 1 would result. In such a severe accident, it is important from the viewpoint of safety that the molted core substance should be kept inside the pressure vessel 1. Description will be made below on operation in the event of occurrence of the severe accident resulting in core melt-down. All the valves can not be operated, the cooling water in the pressure vessel 1 is lost, and the reactor core 3 is melted to drop down onto the bottom head of the pressure vessel 1. Then, the wall temperature of the bottom head of the pressure vessel 1 is increased to open the fuse valve 53 by being melted, the fuse valve 53 being arranged in contact with the outer surface of the bottom head. The cooling water in the pressure suppression pool 14 flows into the lower drywell 13 through the pressure vessel lower portion water flooding pipe 51 to cool the outer surface of the bottom head of the pressure vessel 1. Thereby, the melted reactor core dropped on the bottom head of the pressure vessel 1 is cooled to prevent break of the pressure vessel. Heat from the melted core is transferred to atmospheric air in natural convection inside the ventilation duct 39 placed outside the reactor building through heat conduction in the bottom head of the pressure vessel 1 and heat transfer to the cooling water in the lower drywell plenum 13; blow-down of the generated steam into the pressure suppression pool 14; the heat exchanger 34 or the condensing type heat exchanger 41; and the heat dissipater 43. The heat removal described above can be attained using only the piping and the fuse valve without using any active components such as a pump, a valve and the like, and not limited by an amount of cooling water, and the heat dissipation by atmospheric air can be permanently continued. Therefore, long-term cooling can be performed until the severe accident event is completely settled. According to the first embodiment of the present invention decay heat of the reactor primary system can be removed without using any active components such as a pump and the like and without flowing out the cooling water of the reactor primary system into the containment vessel, it is possible to provide a nuclear reactor which is high in reliability, high in economic feasibility and high in safety. According to the second embodiment of the present invention pressure in the containment vessel can be decreased at an accident of pipe rupture of the reactor primary system, it is possible to provide a nuclear reactor which is high in safety due to reduction of the pressure and high in economic feasibility capable of optimizing design of the containment vessel. According to the third embodiment of the present invention the cooling water in the pressure suppression pool can be injected into the pressure vessel without using any active components such as a pump and the like, it is possible to provide a nuclear reactor which is high in safety and in economic feasibility by improving reliability of the safety components and by simplifying the components. According to the fourth embodiment of the present invention the bottom plenum can be flooded with cooling water and cooled from the outside, it is possible to provide a nuclear reactor which is high in safety by improving reliability of the safety components and by taking the severe accident into consideration, and high in economic feasibility by simplifying the components. According to the fifth embodiment of the present invention cooling water is injected into the pressure vessel using natural phenomena without using any active components such as a valve and the like and heat generated in the reactor core can be released outside the reactor building, it is possible to provide a nuclear reactor which is high in safety by improving reliability of the safety components for reactor core cooling at occurrence of an accident and by improving the long-term cooling performance. According to the sixth embodiment of the present invention it is possible to provide a nuclear reactor which is high in safety, in reliability and in economic feasibility. According to the seventh embodiment of the present invention decay heat of the reactor primary system can be removed without using any active components such as a pump and the like and without flowing out the cooling water of the reactor primary system into the containment vessel, it is possible to provide a nuclear reactor which is high in reliability, high in economic feasibility and high in safety. According to the eighth embodiment of the present invention, high efficient heat exchange can be performed by condensing and liquid film heat transfer in the primary system and by boiling heat transfer in the secondary system, it is possible to provide a nuclear reactor which is high in economic feasibility. According to the ninth embodiment of the present invention the output power of the nuclear reactor can be increased by increasing the total heat transfer area of the heat exchanger, it is possible to provide a nuclear reactor which is high in economic feasibility. According to the tenth embodiment of the present invention the natural circulation flow rate through the reactor core can be increased, it is possible to provide a nuclear reactor which is high in thermal efficiency and in economic feasibility. According to the eleventh embodiment of the present invention the output power of the nuclear reactor can be increased by increasing the total heat transfer area of the heat exchanger, it is possible to provide a nuclear reactor which is high in economic feasibility. According to the twelfth embodiment of the present invention the moisture content in the secondary steam can be reduced to improve the thermal efficiency, it is possible to provide a nuclear reactor which is high in economic feasibility. According to the thirteenth embodiment of the present invention the moisture content in the secondary steam can be further reduced to improve the thermal efficiency, it is possible to provide a nuclear reactor which is high in economic feasibility. According to the fourteenth embodiment of the present invention the heat transfer performance of the heat exchanger can be improved by steam separation of the primary system cooling water to reduce the moisture content in the secondary steam and accordingly to improve the thermal efficiency, it is possible to provide a nuclear reactor which is high in economic feasibility. According to the fifteenth embodiment of the present invention occurrence of an event involving loss of primary cooling water can be prevented, and since cooling of the primary cooling water can be continued and loss of the secondary system cooling water can be prevented even at occurring of pipe rupture in the secondary system piping, it is possible to provide a nuclear reactor which is high in safety. According to the sixteenth embodiment of the present invention, decay heat of the reactor primary system can be removed without using any active components such as a pump and the like and without flowing out the cooling water of the reactor primary system into the containment vessel, it is possible to provide a nuclear reactor which is high in reliability, high in economic feasibility and high in safety. According to the seventeenth embodiment of the present invention, the heat transfer performance of the heat exchanger can be improved by steam separation of the primary system cooling water to reduce the moisture content in the secondary steam and accordingly to improve the thermal efficiency, it is possible to provide a nuclear reactor which is high in economic feasibility.
	summary
047132132	claims	1. A nuclear reactor plant comprising a gas cooled small high temperature reactor, housed in a steel pressure vessel, the reactor having a core containing a pile of spherical fuel elements and traversed from bottom to top by a flow of primary helium gas, a heat utilization system arranged in the flow of cooling gas and installed above the small high temperature reactor in the reactor pressure vessel and which is followed in line preferably by two circulating blowers connected in parallel; the heat utilization system comprising a He/He heat exchanger in which the primary helium transfers its heat to secondary helium circulating in an intermediate circulation loop; the He/He heat exchanger comprises an inner annular coil bundle, extending to a hot gas collector chamber located above the reactor core and exposed from below to hot primary helium cooling gas and an outer coil bundle arranged concentrically and connected successively where the inner coil bundle is of greater length than the outer coil bundle and only the inner coil bundle extends to the hot gas collector chamber; at least one decay heat exchanger installed in the steel pressure vessel; the decay heat exchanger is arranged immediately following the He/He heat exchanger in the direction of flow, and is constantly traversed by the entire flow of cooling gas. 2. A nuclear reactor plant according to claim 1 wherein the steel pressure vessel comprises a drawn-in center part and a re-expanding upper part, with the latter containing the outer coil bundle and the heat exchanger. 3. A nuclear reactor plant according to claim 1, wherein the direction of the flow of primary helium is reversed at an upper end of the inner coil bundle and that the outer coil bundle is exposed to the flow of cooling gas from above. 4. A nuclear reactor plant according to claim 1, wherein the decay heat exchanger has an annular configuration and surrounds the inner coil bundle concentrically below the outer coil bundle. 5. A nuclear plant according to claim 4, wherein the steel pressure vessel comprises a drawn-in center part and a re-expanding upper part with the latter containing the outer coil bundle and the heat exchanger. 6. A nuclear reactor plant according to claim 1, wherein circulating blowers are arranged under the steel pressure vessel. 7. A nuclear plant according to claim 1, wherein the steel pressure vessel has a drawn-in center part and the two circulating blowers are mounted laterally on the steel pressure vessel in an area of the drawn-in center part and shielded against the steel pressure vessel. 8. A nuclear reactor plant according to claim 5, further comprising on the upper part of the pressure vessel, between the decay heat exchanger and the outer coil bundle several connection fittings for inletting of the secondary helium mounted laterally, said fittings leading to a ring header which is connected by the lines with heater surface tubes of the outer coil bundle. 9. A nuclear reactor plant according to claim 8, further comprising a further ring header above the outer coil bundle provided for collecting the secondary helium, means for conducting the secondary helium connecting the ring header of the lines with heater surface tubes of the inner coil bundle. 10. A nuclear reactor plant according to claim 9, further comprising a tube arranged in the center of the inner coil bundle for return of the heated secondary helium, said tube being connected to the inner coil bundle by an outlet header under the inner coil bundle. 11. A nuclear reactor plant according to claim 4, further comprising on the upper part of the pressure vessel, between the decay heat exchanger and the outer coil bundle several connection fittings for inletting of the secondary helium mounted laterally, said fittings leading to a ring header which is connected by the lines with heater surface tubes of the outer coil bundle. 12. A nuclear reactor plant according to claim 11, further comprising a further ring header above the outer coil bundle provided for collecting the secondary helium, means for conducting the secondary helium the ring header connecting of the lines with heater surface tubes of the inner coil bundle. 13. A nuclear reactor plant according to claim 12, further comprising a tube is arranged in the center of the inner coil bundle for return of the heated secondary helium, said tube being connected to the inner coil bundle by an outlet header under the inner coil bundle. 14. A nuclear reactor plant according to claim 3, further comprising on the upper part of the pressure vessel, between the decay heat exchanger and the outer coil bundle connection fittings for inletting of the secondary helium mounted laterally, said fittings leading to a ring header which is connected by the lines with heater surface tubes of the outer coil bundle. 15. A nuclear reactor plant according to claim 14, further comprising a a further ring header above the outer coil bundle provided for collecting the secondary helium, means for conducting the secondary helium the ring header connecting with heater surface tubes of the inner coil bundle. 16. A nuclear reactor plant according to claim 15, further comprising a tube arranged in the center of the inner coil bundle for return of the heated secondary helium, said tube being connected to the inner coil bundle by an outlet header under the inner coil bundle. 17. A nuclear plant according to claim 2, wherein the two circulating blowers are mounted laterally on the steel pressure vessel in an area of the drawn-in center part and shielded against the steel pressure vessel. 18. A nuclear reactor plant according to claim 17, further comprising on the upper part of the pressure vessel, between the decay heat exchanger and the outer coil bundle several connection fittings for inletting of the secondary helium mounted laterally, said fittings leading to a ring header which is connected by the lines with heater surface tubes of the outer coil bundle. 19. A nuclear reactor plant according to claim 18, further comprising a tube arranged in the center of the inner coil bundle for return of the heated secondary helium, said tube being connected to the inner coil bundle by an outlet header under the inner coil bundle.
	description	A WJP method (preventive maintenance method) for a vertical weld portion (or line) of a core shroud (hereinafter referred to as xe2x80x9cshroudxe2x80x9d) in a boiling water reactor (BWR) according to the first embodiment of the present invention is explained by FIG. 1. In this embodiment, an object of the WJP is the vertical weld portion on an outer surface of the shroud. The vertical weld portion is one of narrow space portions in a RPV (reactor pressure vessel). FIG. 1A shows a schematic longitudinal sectional view of the RPV in a state that a top head of the RPV, a steam drier and a shroud head are removed from the RPV. In this state, the RPV 13 is filled with core water 22 and riser pipes 24, jet pumps 25, core cooling pipes 27, etc. are mounted in an annulus portion (a narrow space portion) between the shroud 23 and the RPV 13. In some cases, the vertical weld portion 17 of the shroud 23 is located near the riser pipe 24, and a distance (a spatial width) between the vertical weld portion 17 and the riser pipe 24 is as narrow as about a few tens (20 to 30) mm. In a case that the spatial width is narrow like this, it is impossible to direct a nozzle 4 to the vertical weld portion 17 and to discharge a water jet (hereinafter referred to as xe2x80x9cjetxe2x80x9d) 3 from the nozzle 4. FIG. 1A shows also an ICM housing mounted in a bottom head 26 of the RPV 13. FIG. 1B is a schematic configuration view which shows a state that the WJP method of the present invention is applied to the vertical weld portion on the outer surface of the shroud. In FIG. 1B, the riser pipe 24 is not shown for simplicity. As shown in FIG. 1B, the nozzle 4 is inserted substantially in parallel to the outer surface of the shroud 23 by moving a lifting means 6 using, for example, a fuel exchanger assisting hoist (not shown). Pressurized water flows through a hole in the nozzle 4 and is discharged downward from an opening of the nozzle 4 as a jet 3. When the jet 3 is discharged from the opening, cavitation bubbles 2a are generated. This jet 3 containing cavitation bubbles 2a collides with (or impinges on) a plane surface (hereinafter referred to as xe2x80x9ca collision surfacexe2x80x9d) of a baffle body 5a provided near the vertical weld portion 17. The jet 3 changes direction and velocity of its flow by the collision with the collision surface of the baffle body 5a, and collides with the vertical weld portion 17 as a collision jet 9a. That is, the baffle body 5a is a deflector of the jet 3. Although it is omitted in FIG. 1B, practically, a relative position between the nozzle 4 and the collision surface of the baffle body 5a is maintained by a support. A distance between an end of the nozzle 4 and the collision surface 50 of the baffle body 5a is defined as a collision distance L as shown in FIG. 1C. Strictly, the collision distance L is a distance in a central axis 3a passing through the opening of the nozzle 4. In this embodiment, the collision distance L is set at most 100 times (preferably at most 50 times) as large as a hole diameter of the nozzle 4. This hole diameter means a substantial diameter of the hole in the nozzle 4. By arranging the nozzle 4 and the baffle body 5a so as to meet the above condition, the jet 3 collides with the collision surface 50 before fine cavitation bubbles contained in the jet 3 become large. Therefore, since the amount (a ratio) of the cavitation bubbles collapsed by the collision with the collision surface 50 is reduced and the jet 3 collides with the collision surface 50 before its velocity becomes low, the collision jet 9a including a strong vortex flow and a strong separation flow is generated. Accordingly, the fine cavitation bubbles, which are not collapsed by the collision with the collision surface 50, grow in the collision jet 9a and collapse at the vertical weld portion 17 with high collapse pressures, thereby a tensile residual stress of the vertical weld portion 17 can be reduced effectively. If the collision distance L is set more than 100 times as large as the hole diameter of the nozzle 4, the amount (the ratio) of the cavitation bubbles collapsed by the collision with the collision surface 50 becomes large and the velocity of the jet 3 becomes low. Therefore, the amount (the ratio) of the cavitation bubbles contained in the collision jet 9a is reduced and an improvement effect of the residual stress decreases. As shown in FIG. 1C, an angle formed the central axis 3a passing through the opening of the nozzle 4 and the collision surface 50 is defined as a collision angle xcex1. Strictly, the collision angle a is a lower (smaller) angle of two angles formed the central axis 3a and the collision surface 50 on a plane 51 including both the central axis 3a and a perpendicular line 50a of the collision surface 50, the perpendicular line 50a passing through an intersection point where the central axis 3a crosses the collision surface 50. The collision angle a is an acute angle except a case that the central axis 3a crosses perpendicularly the collision surface 50. The collision angle xcex1 is needed to be at least 10xc2x0. When the jet 3 collides with the collision surface 50, not only the collision jet 9a flowing toward the vertical weld portion 17 but also, for example, a collision jet 9b flowing opposite to the vertical weld portion 17 is generated. If the collision angle xcex1 is set about 10xc2x0, since the collision surface has a steep slope (incline) to the vertical weld portion 17, a rate of the collision jet 9a can be higher and a rate of the collision jet 9b can be lower in comparison with a case of a xcex1 less than 10xc2x0. In this case, however, the vortex flow and the separation flow in the collision jet 9a are not so strong because the water-hammering effect on the collision surface 50 is still weak. Therefore, a long period of time for discharging the jet 3 is needed to attain a desired effect of improving the residual stress. In this embodiment, the collision angle xcex1 is set in a range of 40xc2x0 to 90xc2x0 (preferably in a range of 60xc2x0 to 90xc2x0). In this case, since the water-hammering effect on the collision surface 50 becomes strong, the strong vortex flow and the strong separation flow can be generated in the collision jet 9a. Accordingly, it is possible to impinge the collision jet 9a containing the cavitation bubbles with the high collapse pressures on the vertical weld portion 17, and also attain the desired effect of improving the residual stress more effectively. According to this embodiment, it is easy to indirectly impinge the jet 3 on the vertical weld portion 17 without directing the nozzle 4 to the vertical weld portion 17. When the jet 3 collides with the collision surface 50, part of cavitation bubbles 2a contained in the jet 3 collapse due to an increase of a fluid pressure caused by the water-hammering effect. But the remaining cavitation bubbles, which do not collapse on the collision surface 50, grow to the cavitation bubbles with the high collapse pressures in the collision jet 9a including the strong vortex flow and the strong separation flow. In the collision jet 9a, in addition to the above mentioned growth of the remaining cavitation bubbles, new cavitation bubbles are also generated and then grow. As a result, the collapse pressure of the collision jet 9a on the vertical weld portion 17 becomes higher, and it is possible to attain the effect of improving significantly the residual stress of the vertical weld portion 17. FIG. 1D shows another example of the baffle body which is used for changing the direction of the flow of the jet 3 in FIG. 1B. This baffle body has a curved surface 5d as the collision surface and jet guids 5dxe2x80x2 which are provided at both sides of the curved surface 5d. In a case of using this baffle body, the strong vortex flow and the strong separation flow are generated in the collision jet 9a, and the collision jet 9a containing the cavitation bubbles with the high collapse pressures can collide with (impinge on) the vertical weld portion 17. Further, it is possible to reduce effectively the rate (amount) of collision jet except the collision jet 9a flowing toward the weld portion 17. One example of a nozzle head, which can discharge a collision jet to almost one direction, according to the present invention is explained by FIG. 2. FIG. 2 shows a schematic configuration view of a nozzle head 15a which is a one-sided discharging type and has a flow baffle 5 with an opening at one side. Hereinafter, this flow baffle is referred to as xe2x80x9ca one-sided opening type flow bafflexe2x80x9d. This flow baffle 5 is formed into a cylindrical shape and has a square-shaped opening 5b which is formed by cutting out a circumferential part near one end portion of the cylinder. A baffle body 5a is removably engaged with the one end portion of the flow baffle 5 at a position adjacent to the opening 5b in such a manner that a collision jet 9a passing through the opening 5b collides with a surface to be treated. The nozzle head 15a is constructed by engaging removably a nozzle 4 with the other end portion of the flow baffle 5. Since the baffle body 5a is removably engaged with the flow baffle 5, when the baffle body 5a is worn, it can be easily replaced with a new one. Therefore, reliability of execution of WJP can be maintained. In this nozzle head 15a, a collision distance and a collision angle are set in the above-mentioned range. In FIG. 2, the jet 3 collides with the collision surface of the baffle body 5a to change its flow direction, and the collision jet 9a directly collides with the surface to be treated. A collision jet 9b flowing toward direction in which the opening 5b is not provided, changes its flow direction toward the opening 5b by making a second collision with an inner wall of the flow baffle 5, and are discharged from the opening 5b so as to make a third collision with the surface to be treated. In this case, cavitation bubbles in the collision jet grow more largely by this second collision, and the collision jet can restrictively collide with the surface to be treated. Further, by making fine irregularities on the collision surface of the baffle body 5a, the cavitation bubbles grow largely by the collision with the collision surface having the fine irregularities. This growth of the cavitation bubbles can make a strong peening effect (a strong effect of improving the residual stress) in cooperation with the above mentioned repeated collision. In FIG. 2, it is possible to replace the cylindrical flow baffle with a square pipe flow baffle. It is also possible to replace the plane collision surface with a curved surface as shown in FIG. 1C. FIG. 3 shows one example of an improvement effect of the residual stress by using the one-sided discharging type nozzle head 15a shown in FIG. 2. The nozzle having an outer diameter of 30 mm and a hole diameter of 2 mm is used. The baffle body 5a is arranged so as to make the collision distance of 80 mm and the collision angle of 70xc2x0. The one-sided opening type flow baffle 5 has the opening 5b in a half circumferential part. FIG. 3 shows a measurement result of the residual stress on a surface of a strip-shaped (plate-shaped) test piece after executing the WJP to the test piece using this nozzle head 15a. The WJP is executed in a condition that the nozzle head is moving to a longitudinal direction (Y-direction) by keeping a distance between the nozzle head and the surface of the test piece about 5 mm. In FIG. 3, a vertical axis is a relative measurement value of the residual stress, and a horizontal axis is a distance from a center line (Y-axis) of the test piece in a width direction (X-direction). A positive residual stress is a tensile residual stress, and a negative residual stress is a compressive residual stress. The surface of the test piece is subjected to surface grinding so as to have a tensile residual stress of about 400 MPa as an initial residual stress. As shown in FIG. 3, the initial tensile residual stress is improved to the compressive residual stress in a range in which the collision jet collides with the surface of the test piece. The first embodiment, in which the WJP method according to the present invention is applied to the vertical weld portion on the outer surface of the shroud in a BWR plant after at least the first operation cycle, is explained in more detail using FIG. 4 and FIG. 5. A WJP apparatus having the one-sided discharging type nozzle head 15a with the cylindrical flow baffle 5 is used. The collision distance and the collision angle are set in the above-mentioned range, respectively. FIG. 4 is a schematic longitudinal sectional view, which shows a state of the WJP execution, of a surrounding area near the RPV. FIG. 4 also shows the third embodiment in which the WJP method according to the present invention is applied to a horizontal weld portion of an ICM housing. FIG. 5 is a schematic flow chart which shows execution steps of the WJP In the first embodiment. Each stop is explained below according to the flow chart of FIG. 5. (1) Disconnection: A top head of the RPV, a steam drier and a shroud head are removed from the RPV. In this state, the RPV 13 and a reactor well are filled with core water 22. (2) Detection of weld line: A weld line detector (not shown) is lowered and set near an outer surface of the shroud using, for example, a fuel exchanger assisting hoist (hereinafter referred to as xe2x80x9cassisting hoistxe2x80x9d) 21. A vertical weld portion (line) is detected by the weld line detector. (3) Confirmation of access route: While a monitor camera 30 is lowered using, for example, the assisting hoist 21, an access route to the weld line 17, presence or absence of an obstacle to set a WJP main body 29, and the weld line 17 are confirmed by means of a monitor video 31. A spatial distance between a riser pipe 24 and the shroud 23 is measured to confirm that a nozzle head can be inserted into the space. (4) Setting of WJP apparatus: A control panel 20 and a booster pump 18 are disposed on an operation floor. The booster pump 18 is connected to a source water tank (not shown) by means of a water supply hose 19. The booster pump 18 is connected to the WJP main body 29 by means of a high-pressure hose 7. Wiring between these devices is laid out, and these devices are adjusted. (5) Setting of WJP main body: This step has next steps of a) to e). a) Lowering: The WJP main body 29 is lowered by the assisting hoist 21 to a specific height in a space between the shroud 23 and the RPV 13. It is confirmed by the monitor camera 30 and the monitor video 31 that the WJP main body 29 is located in a suitable height. b) Fixing: Upper and lower portions of the WJP main body 29 are fixed on a shroud""s side and a RPV""s side by a support 29a and a support 29b. c) Extending nozzle arm: A nozzle head 15a fixed at a top end of a nozzle arm 33 is inserted between the shroud 23 and the riser pipe 24 by extending forwardly the nozzle arm 33. d) Confirmation of position: A distance between the weld line 17 and the nozzle head 15a and discharging direction are confirmed by the monitor camera 30 and the monitor video 31. e) Trial discharge of jet: A trial discharge of a collision jet 9 is performed to confirm that the collision jet 9 collides with a desired position by the monitor camera 30 and the monitor video 31. It is the last step for setting of the WJP main body 29. (6) Execution of WJP: This step has next steps of a) to c). a) Setting of execution conditions: A discharging pressure and a flow rate of the jet, and a moving speed and a moving range of the nozzle head 15a are set. b) Discharge of jet: The collision jet 9 is discharged and the nozzle head 15a is moved in a vertical direction along the weld line 17 to execute the WJP. This execution state of the WJP is confirmed by the monitor camera 30 and the monitor video 31. In this state, the schematic longitudinal sectional view of the surrounding area near the RPV is shown in FIG. 4A, and a top view of a surrounding area near the WJP main body 29 is shown in FIG. 4B. c) Confirmation of execution of WJP: A state of a surrounding area near the weld line 17 after the execution of the WJP is confirmed by the monitor camera 30 and the monitor video 31 to terminate the execution of the WJP. (7) Withdrawal of WJP main body: This step has next steps of a) to d). a) Folding of nozzle arm: The nozzle arm 33 is folded to be contained in the WJP main body 29. b) Release of main body: The WJP main body 29 fixed between the shroud 23 and the RPV 13 is released. c) Confirmation of preparation for lifting: A termination of preparation for lifting the WJP main body 29 is confirmed by the monitor camera 30 and the monitor video 31. d) Lifting of main body: The WJP main body 29 is lifted by the assisting hoist 21. (8) Withdrawal of WJP apparatus: The connection between the booster pump 18 and the source water tank by the water supply hose 19 and the connection between the booster pomp 18 and the WJP main body 29 by the high pressure hose 7 are released, and the wiring between these devices is removed. The apparatuses such as the WJP main body 29, the control panel 20, the booster pump 18, the high pressure hose 7 and the water supply hose 19 are withdrawn. (9) Withdrawal of monitor camera: The monitor camera 30 is withdrawn. (10) Withdrawal of weld line detector: The weld line detector is withdrawn to terminate the execution of the WJP. (11) Synchronization: The shroud head, the steam drier, and the top head of the RPV are lowered and assembled to be restored. By executing (applying) the WJP with the above steps to the vertical weld portion on the outer surface of the shroud in the RPV filled with the core water, it is possible to collapse cavitation bubbles with high collapse pressures on a surface of the vertical weld portion. Accordingly, the residual stress on the surface of the vertical weld portion can be improved and a damage such as the SCC can be prevented. When the above WJP method is executed during an outage of the BWR plant, since the top head of the RPV, the steam drier and the shroud head are already removed, the execution of the WJP is started from the step (2) and terminated at the step (9). The one-sided discharging type nozzle head 15a can be applied to axial weld lines on both inner and outer surfaces of a weld pipe. Of course, it can be applied to a weld pipe with no weld line. One example of a four-sided discharging type nozzle head according to the present invention is explained by FIG. 6. FIG. 6A shows a schematic configuration view of this nozzle head, and FIG. 6B shows an Axe2x80x94A cross sectional view of FIG. 6A. This nozzle head 15b has a cylindrical flow baffle 5 with four square openings 5b which are arranged symmetrically in a peripheral direction. Each of four supports 5x forming the openings 5b has a square-shaped cross section. A baffle body 5a having a flat collision surface is removably engaged with one end portion of the flow baffle 5 at a position adjacent to the openings 5b. A nozzle 4 is removably and rotatably engaged with the other end portion of the flow baffle 5. A collision angle is about 90xc2x0 and a collision distance is set in the above-mentioned range. Since the baffle body 5a is removably engaged with the flow baffle 5, when the baffle body 5a is worn, it can be easily replaced with a new one. Therefore, reliability of execution of WJP can be maintained. In this nozzle head 15b, a jet 3 having cavitation bubbles collides with the collision surface of the baffle body 5a and is discharged from the four openings 5b as four collision jets 9a. Therefore, it is possible to execute the WJP simultaneously to a plurality of objects to be treated which are disposed opposite to the four openings 5b. In this case, since velocity of the collision jets 9a in an axial direction becomes almost zero, a strong water-hammering effect and a turbulent flow are generated, and an vortex flow and a separation flow generated in the collision jet become strong. In this nozzle head 15b, by making width of each opening 5b wider, the collision jets 9a can be discharged in approximately radial directions. In this case, the nozzle head 15b can make an almost omni-directional discharge which is suitable for executing the WJP to an entire inner surface of a cylinder. Therefore, by discharging the jet from this nozzle head to a peripheral weld portion on an inner surface of such a tube with a small diameter, it is possible to execute the WJP simultaneously to the entire peripheral weld portion without rotating this nozzle head from outside. Also, by increasing the number of the openings 5b, the collision jets 9a can be discharged in approximately radial directions. Further, in this nozzle head 15b, since the openings 5b are made longer in the axial direction, the jet 3 can draw water near the openings 5b. Therefore, since cavitation bubbles contained in the jet 3 can grow largely before the collision with the baffle body 5a, the improvement effect of the residual stress by the collision jet becomes higher. Another example of a four-sided discharging type nozzle head according to the present invention is explained by FIG. 6C. FIG. 6C shows a cross sectional view which corresponds to the Axe2x80x94A cross sectional view of FIG. 6A. In this nozzle head, each of four supports 5x forming the openings 5b has curved sides as shown in FIG. 6C. As a result, the support 5x has an almost parallelogram-shaped cross section. The collision jets 9a become to have velocity components in both a radial direction and a peripheral direction by passing through this openings 5b. That is, the collision jets 9a become a revolving flow. In this nozzle head, since the collision jets 9a become the revolving flow, the collision jets 9a can go around to portions which are not disposed opposite to the openings. Further, the nozzle 4 is not rotated but the flow baffle 5 is rotated on its axis by a reaction force to the revolving flows. Therefore, this nozzle head is more suitable for executing the WJP to the entire inner surface of the cylinder than that shown in FIG. 6B. That is, this nozzle head can make an almost omni-directional discharge of the collision jets. Another example of a four-sided discharging type nozzle head according to the present invention is explained by FIG. 7. FIG. 7A shows a cross sectional view which corresponds to FIG. 6B. FIG. 7B and FIG. 7C show a Bxe2x80x94B cross sectional view and a Cxe2x80x94C cross sectional view of FIG. 7A, respectively. The other elements of this nozzle head are almost the same as FIG. 6A. As shown in FIG. 7A, this nozzle head has a collision surface with four spiral grooves 5c which are symmetrical with respect to an central axis of the collision surface. As shown in FIG. 7C, each groove 5c has a V-shaped cross section. In this nozzle head, the collision jet 9a discharged from the opening is given a velocity component in a peripheral direction by the groove 5c. That is, the collision jets 9a become a revolving flow. As a result, the collision jets 9a can go around to portions which are not disposed opposite to the openings. Therefore, this nozzle head is also suitable for executing the WJP to the entire inner surface of the cylinder. Further, if the spiral grooves 5c are replaced with spiral projections, the same effect can be attained. In FIG. 7A, the spiral grooves 5c are originated from positions which are separated from the central axis of the collision surface. If the spiral grooves 5c are originated from the central axis of the collision surface, since vortex flows and separation flows contained in the collision jets 9a become stronger, the collision jets can become collision jets containing cavitation bubbles with high collapse pressures. Therefore, higher improvement effect of the residual stress can be attained. Further, by combining the spiral grooves 5c with the supports 5x shown in FIG. 6C, the peripheral velocity component of the collision jet 9a becomes higher and the rotation speed of the flow baffle 5 on its axis also becomes higher. Therefore, the improvement effect of the residual stress can be attained more effectively. Another example of a four-sided discharging type nozzle head according to the present invention is explained by FIG. 7D. FIG. 7D shows a longitudinal sectional view which corresponds to FIG. 7B. The other elements of this nozzle head are almost the same as FIG. 6A. This nozzle head has a recessed baffle body 5a which has a recess with a concave cross section as the collision surface. The recess is in shape of cone with an apex angle xcex2 of at least 90xc2x0 (preferably at least 120xc2x0) in a longitudinal cross section thereof. When a jet 3 collides with the collision surface, velocity of the jet 3 in a collision direction (a downward direction in FIG. 7D) becomes zero on the collision surface, and then the jet 3 changes to a collision jet 9a with a velocity component in direction (an upward direction in FIG. 7D) opposed to that of the jet 3. Since a change in velocity from the jet 3 to the collision jet 9 becomes large by setting the apex angle xcex2 in the above range, a water-hammering effect occurs strongly on the collision surface. Therefore, part of cavitation bubbles collapse strongly on the collision surface. The remaining cavitation bubbles, which are not collapsed on the collision surface, grow in a strong vortex flow and a strong separation flow included in the collision flow 9a, and are discharged. Also, in this nozzle head, by forming spiral grooves (or spiral projections) as shown in FIG. 7A on the collision surface, it is possible to give a revolving flow to the collision jet 9a and also generate the vortex flow and the separation flow more strongly. As a result, an improvement effect of the residual stress which is high and almost uniform in the peripheral direction can be obtained. As a modification of FIG. 7D, the collision surface can be formed into a projecting surface (shape). In this case, the top of the projecting surface breaks a central flow in the jet 3 and generates cavitation bubbles. Further, it becomes easy to form grooves (or projections) like FIG. 7A on the collision surface by machining. The second embodiment, in which the WJP method according to the present invention is applied to a weld portion of a water-level measuring nozzle in a BWR, is explained using FIG. 8. FIG. 8 is a longitudinal sectional view which shows a state that a nozzle head 15b is set in a water-level measuring nozzle 35. An object of the WJP in this embodiment is a weld portion 38 between a nozzle 36 and a safe end 37 in the water-level measuring nozzle 35 mounted in a RPV 13. The nozzle head 15b shown in FIG. 7D is used in this embodiment. A central flow (a flow near a central axis) in a jet 3 changes its flow direction by a collision with a central portion of a recessed surface (collision surface) and then flows along the recessed surface, thereby a strong turbulent flow is generated by interference between the direction-changed flow and an outer flow in the jet 3. A collision jet generated like this flows toward the RPV 13 (a right side in FIG. 8) in the water-level measuring nozzle 35, and is finally discharged into the RPV 13 because a leading end of the water-level measuring nozzle 35 is closed with a valve 37a. An apparatus used for execution of the WJP to the weld portion 38 in the water-level measuring nozzle 35 is explained using FIG. 9. This apparatus has a nozzle head drive unit 39 for moving the nozzle head 15b to an object to be treated, a frame 40 for supporting the nozzle head drive unit 39 at a level of the water-level measuring nozzle 35, a high-pressure hose 42 and a booster pump 43 for supplying pressurized water to a nozzle 4, a water supply hose 44 for supplying water to the booster pump 43, and a control panel 45 for controlling the nozzle head drive unit 39 and the booster pump 43. The WJP is executed using the above apparatus in accordance with the following steps. (1) Disconnection: A top head of the RPV, a steam drier, a shroud head and fuel assemblies are removed from the RPV. In this state, the RPV 13 and a reactor well are filled with core water 22. (2) Setting of nozzle head drive unit: The nozzle head drive unit 39 is mounted on the frame 40. The nozzle head drive unit 39 is lowered in the frame 40 by an assisting hoist 21, and is set at a position of the water-level measuring nozzle 35. (3) Preparation for execution of WJP: This step has next steps of a) to c). a) Setting of WJP apparatus: The nozzle head 15b mounted at a top end of the nozzle head drive unit 39 is inserted in the water-level measuring nozzle 35. The control panel 45 and the booster pump 43 are disposed on an operation floor. The booster pump 43 is connected to a source water tank 46 by the water-supply hose 44. The booster pump 43 is connected to a WJP main body by the high-pressure hose 42. Wiring between these devices are laid out, and these devices are adjusted. b) Setting of execution conditions: A flow rate and a discharging period (time) of the jet, a moving speed and a moving range of the nozzle head in an axial direction, and a turning speed and a turning range of the nozzle head in a peripheral direction are set. c) Confirmation of operation of apparatus: The nozzle head 15b is moved according to the setting conditions in a head 15b is moved according to the setting conditions in a state in which the jet is not discharged, to confirm whether or not the execution range is suitable, the nozzle head 15b is smoothly moved, and the like. (4) Execution of WJP: A jet is discharged to start the execution of the WJP. (5) Confirmation of execution of WJP: This step has next steps of a) to b). a) Removal of nozzle head: The nozzle head 15b is removed from the water-level measuring nozzle 35. b) Confirmation of execution of WJP: A monitor camera 47 is inserted in the water-level measuring nozzle 35. It is confirmed by a monitor TV 48 that the WJP is suitably executed. The suitably executed state is recorded in a monitor video 49. (6) Withdrawal of WJP apparatus: This step has next steps of a) to b). a) Withdrawal of monitor camera: The monitor camera 48 is removed from the water-level measuring nozzle 35 to be withdrawn. b) Withdrawal of WJP apparatus: Piping and wiring between the above devices are removed. The devices, pipes for piping, and wires for wiring are withdrawn. (7) Synchronization: The fuel assemblies, the shroud head, the steam drier, and the top head of the RPV are lowered and assembled to be restored. By executing (applying) the WJP with the above steps to the weld portion of the water-level measuring nozzle in the RPV filled with the core water, it is possible to collapse cavitation bubbles with high collapse pressures on a surface of the weld portion. Accordingly, the residual stress on the surface of the weld portion can be improved and a damage such as the SCC can be prevented. Another example of a four-sided discharging type nozzle head according to the present invention is explained by FIG. 10. FIG. 10 shows a schematic configuration view of this nozzle head. This nozzle head has a turning vane 5d adjacent to the baffle body 5a on an opposite side to the nozzle. The turning vane 5d and the baffle body 5a have the same central axis. That is, this nozzle head has the flow baffle 5 with the turning vane 5d. The other elements of this nozzle head are almost the same as FIG. 6A. In this case, the turning vane 5d is turned by the collision jet which changed its flow direction by the collision with an object to be treated, and this rotation of the turning vane 5d assists a rotation of the baffle body 5a on its axis. The third embodiment, in which the WJP method according to the present invention is applied to an inner surface of a horizontal weld portion (or line) of an ICM housing in a BWR, is explained using FIG. 11. FIG. 11 is a schematic configuration view which shows a state that a nozzle head 15c with a back-flow obstructive plate 10 is set at the inner surface of the horizontal weld portion 17a of the ICM housing 1. As shown in FIG. 4A, the TCM housing 1 pierces a bottom head 26 of the RPV 13 and is fixed to the bottom head 26. The nozzle head 15c corresponds to the nozzle head 15b (shown in FIG. 8) to which the back-flow obstructive plate 10 is added on a nozzle side. Since the nozzle head 15c has the back-flow obstructive plate 10, a sealing portion located at a lower end of the ICM housing 1 is protected for sealing water. In this nozzle head 15c, the collision jet, which changed its flow direction opposite to an initial flow direction of the jet 3 by the collision with the buffle body 5a, can change (be repelled) its flow direction to the initial flow direction by a collision with the back-flow obstructive plate 10. Interference between this repelled collision jet and the initial collision jet makes a turbulent flow, and this turbulent flow can make the peening effect higher. As shown in FIG. 11, an apparatus used for execution of the WJP to the horizontal weld portion 17a of the TCM housing 1 has a lifting shaft 6a with a lifting guide 14 mounting the nozzle head 15c, a nozzle head drive unit 16 having a nozzle rotating means 16a provided at a lower end of the lifting shaft 6a and a nozzle lifter 16b, a high-pressure hose 7 and a booster pump 18 for supplying pressurized water to a nozzle 4, a water supply hose 19 for supplying water to the booster pump 18, and a control panel 20 for controlling the nozzle head drive unit 16 and the booster pump 18. The WJP is executed using the above apparatus in accordance with the following steps. (1) Disconnection: A top head of the RPV, a steam drier, a shroud head, fuel assemblies and control rods are removed from the RPV. In this state, the RPV 13 and a reactor well are filled with core water 22. (2) Water sealing for upper end of ICM guide tube: The upper end of the TCM guide tube above the ICM housing 1 shown in FIG. 1A is plugged for sealing water. (3) Removal of ICM detector: The ICM detector (not shown) contained in the ICM guide tube is removed from the lower end of the ICM housing 1. (4) Confirmation of welding position: An ultrasonic sensor (not shown) or the like is inserted from the lower end of the ICM housing 1 to confirm a position of the horizontal weld portion 17a and an execution range of the WJP. (5) Preparation for execution of WJP: This step has next steps of a) to e). a) Setting of WJP apparatus: A nozzle drive shaft 16c mounting the nozzle head 15c at the leading end is inserted in the ICM housing 1. The nozzle head drive unit 16, control panel 20 and the booster pump 18 are disposed as shown in FIG. 4A. The booster pump 18 is connected to a source water tank (not shown) by the water supply hose 19. The booster pump 18 is connected to a WJP main body by the high-pressure hose 7. Wiring between these devices is laid out, and these devices are adjusted. b) Setting of execution conditions: A flow rate and a discharging period (time) of the jet, a moving speed and a moving range of the nozzle head in an axial direction, and a turning speed and a turning range of the nozzle head in a peripheral direction are set. c) Confirmation of operation of apparatus: The nozzle head 15c is moved according to the setting conditions in a state in which the jet is not discharged, to confirm whether or not the execution range is suitable, the nozzle head 15c is smoothly moved, and the like. d) Release of sealing of upper end of ICM guide tube: The plugging of the upper end of the ICM guide tube is released. e) Trial discharge of jet: The trial discharge of the jet 3 is performed for conforming looseness of pipes, a vibrational state, and the like. In this way, the setting of the WJP apparatus is terminated. (6) Execution of WJP: The jet 3 is discharged to start execution of the WJP to the horizontal weld portion 17a. (7) Confirmation of execution of WJP: This stop has next steps of a) to d). a) sealing for upper end of ICM guide tube: The upper end of the ICM guide tube is plugged for sealing water. b) Removal of nozzle drive shaft: The nozzle drive shaft 16c is removed from the ICM housing 1. c) Setting of monitor camera: The monitor camera 30 is inserted in the ICM housing 1 and is set in co-operation with the monitor video 31. d) Confirmation of execution of WJP: It is confirmed that the WJP is suitably executed by the monitor camera 30. (8) Withdrawal of WJP apparatus: This step has next steps of a) to b). a) Withdrawal of monitor camera: The monitor camera 30 is removed from the ICM housing 1 to be withdrawn. b) Withdrawal of WJP apparatus: The wiring and piping between the above devices are removed, and the devices, pipes for piping, and wires for wiring are withdrawn. (9) Mounting of ICM detector: The ICM detector is inserted from the lower end of the ICM housing 1 to be mounted. (10) Release of sealing of upper end of ICM guide tube: The plugging of the upper end of the ICM guide tube is released. In this way, the execution of the WJP is terminated. (11) Synchronization: The fuel assemblies, the control rods, the shroud head, the steam drier and the top head of the RPV are lowered and assembled to be restored. By executing (applying) the WJP with the above steps to the horizontal weld portion of the ICM housing in the RPV filled with the core water, it is possible to collapse cavitation bubbles with high collapse pressures on a surface of the horizontal weld portion. Accordingly, the residual stress on the surface of the horizontal weld portion can be improved and a damage such as the SCC can be prevented. In this embodiment, the nozzle head 15c with the back-flow obstructive plate 10 is used. However, the nozzle head 15a and 15b shown in FIGS. 2 and 6 are also can be used in the above steps. The fourth embodiment, in which the WJP method according to the present invention is applied to an inner surface of a horizontal weld portion of an ICM housing in a BWR, is explained using FIG. 12. In the third embodiment, the nozzle head is inserted from the lower end of the ICM housing, however, in this embodiment, the nozzle head is inserted from the upper end of the ICM housing. In this embodimen. The nozzle head 15b shown in FIG. 8 is used. FIG. 12 is a schematically constructional view which shows a state that the nozzle head 15b is set at the inner surface of the horizontal weld portion 17a of the ICM housing 1. Executing steps of WJP according to this embodiment is explained below. (1) Disconnection: A top head of the RPV, a steam drier, a shroud head, fuel assemblies and control rods are removed from the RPV. In this state, the RPV 13 and a reactor well are filled with core water 22. (2) Water sealing for upper end of ICM guide tube and removal of ICM detector: This step has next steps of a) to b). a) The upper end of the ICM guide tube above the ICM housing 1 shown in FIG. 1A is plugged for sealing water. In FIG. 12, 34 is a core support. b) The ICM detector (not shown) contained in the ICM guide tube 1a is removed from the lower end of the ICM housing 1. (3) Water sealing for lower end of ICM housing and release of sealing for upper end of ICM guide tube: This step has next steps of a) to b). a) A closing flange 32 is mounted at the lower end of the ICM housing 1 for sealing water. b) The plugging of the upper end of the ICM guide tube is released. (4) Confirmation of welding position: An ultrasonic sensor (not shown) or the like is inserted from the upper end of the ICM guide tube 1a to confirm a position of the horizontal weld portion 17a and an execution range of the WJP. (5) Preparation for execution of WJP: This step has next steps of a) to d). a) Setting of WJP apparatus: A lifting shaft 6a mounting the nozzle head 15b at the leading end is inserted in the ICM housing 1 from an upper side. A nozzle head drive unit 16, a control panel 20 and a booster pump 18 are disposed. The booster pump 18 is connected to a source water tank (not shown) by the water supply hose 19. The booster pump 18 is connected to a WJP main body by a high-pressure hose 7. Wiring between these devices is laid out, and these devices are adjusted. The arrangement of these devices in this case are substantially the same as those shown in FIG. 4A. Therefore, the explanation thereof is omitted. b) Setting of execution conditions: A flow rate and a discharging period (time) of the jet, a moving speed and a moving range of the nozzle head in an axial direction, and a turning speed and a turning range of the nozzle head in a peripheral direction are set. c) Confirmation of operation of apparatus: The nozzle head 15b is moved according to the setting conditions in a state in which the jet is not discharged, to confirm whether or not the execution range is suitable, the nozzle head 15c is smoothly moved, and the like. d) Trial discharge of jet: The trial discharge of the jet 3 is performed for conforming looseness of pipes, a vibrational state, and the like. In this way, the setting of the WJP apparatus is terminated. (6) Execution of WJP: The jet 3 is discharged to start execution of the WJP to the horizontal weld portion 17a. (7) Confirmation of execution of WJP: This step has next steps of a) to c). a) Removal of nozzle drive shaft: The nozzle drive shaft (not shown) is removed from the ICM housing 1. b) Setting of monitor camera: The monitor camera is inserted in the ICM housing 1 and is set in co-operation with the monitor video. c) Confirmation of execution of WJP: It is confirmed that the WJP is suitably executed by the monitor camera. (8) Withdrawal of WJP apparatus: This step has next steps of a) to b). a) Withdrawal of monitor camera: The monitor camera is removed from the ICM housing 1 to be withdrawn. b) Withdrawal of WJP apparatus: The wiring and piping between the above devices are removed, and the devices, pipes for piping, and wires for wiring are withdrawn. (9) Water sealing for upper end of ICM guide tube, mounting of ICM detector, and release of sealing of upper end of ICM guide tube: This step has next steps of a) to c). a) The upper end of the ICM guide tube 1a is plugged for sealing water. b) The closing flange 32 at the lower end of the ICM housing 1 is removed, and the ICM detector is inserted to be mounted. c) The plugging of the upper end of the ICM guide tube 1a is released. In this way, the execution of the WJP is terminated. (10) Synchronization: The fuel assemblies, the control rods, the shroud head, the steam drier and the top head of the RPV are lowered and assembled to be restored. In this embodiment, since the nozzle head 15b having the baffle body 5a with the recessed surface (collision surface) is used, a central flow in the jet 3 changes its flow direction by the collision with a central portion of the recessed surface and then flows along the recessed surface, thereby a strong turbulent flow is generated by interference between the direction-changed flow and an outer flow in the jet 3. A collision jet generated like this flows upward in the ICM housing 1, and is finally discharged into the RPV 13 because the closing flange 32 is mounted at the lower end of the ICM housing 1. Since the collapse pressures of the cavitation bubbles become higher by a strong turbulent flow generated near the recessed surface, a high improvement effect of the residual stress can be obtained. Accordingly, in this embodiment, the residual stress on the surface of the horizontal weld portion of the ICM housing can be improved and damage such as the SCC can be prevented like in the third embodiment. FIG. 13 shows one example of the improvement effect of the residual stress by using the four-sided discharging type nozzle head 15b shown in FIG. 6A. The nozzle 4 having an outer diameter of 30 mm and a hole diameter of 2 mm is used. The baffle body 5a is arranged so as to make the collision distance of 80 mm and the collision angle of 90xc2x0. FIG. 13 shows a measurement result of the residual stress on an inner surface of a test tube with an inner diameter of 38 mm after executing the WJP to the inner surface of the test tube using this nozzle head 15b. The WJP is executed in a condition that the nozzle head is moving to an axial direction (Z-direction) of the test tube. In FIG. 13, a vertical axis is a relative measurement value of the residual stress, and a horizontal axis is a distance from a center position in an executing region of WJP in the Z-direction. A positive residual stress is a tensile residual stress, and a negative residual stress is a compressive residual stress. The test tube is divided into three pieces, and its surface is subjected to surface grinding so as to have a tensile residual stress of about 400 MPa as an initial residual stress. As shown in FIG. 13, the initial tensile residual stress is improved to the compressive residual stress by executing the WJP. Since it is known that no SCC and no fatigue fracture occur under the compressive stress, it is possible to prevent the SCC and the fatigue fracture by applying the above-mentioned WJP in accordance with the present invention. While the WJP methods according to the present invention are applied to the structural members in the RPV in the above-mentioned embodiments, objects applied by these WJP methods are not limited in this. That is, these WJP methods can be applied to tubes in a nuclear plant, general industrial machines and ships.
	summary
	claims	1. A scintillator single crystal expressed by(PrxLu1−x)3Al5O12 where 0.002≦x≦0.02; andthe scintillator single crystal emits fluorescence having a wavelength of 200 to 350 nm when excited by gamma ray and having a decay time at room temperature of 1 to 50 nsec, andwherein the scintillator single crystal is a crystal formed by Czochralski method. 2. The scintillator single crystal according to claim 1, wherein the scintillator single crystal has an emission peak at a wavelength of 200 to 350 nm. 3. The scintillator single crystal according to claim 1, wherein the scintillator single crystal has an emission peak at a wavelength of about 300 nm. 4. The scintillator single crystal according to claim 1,wherein a fluorescence decay time of the scintillator single crystal is shorter than 20 nsec. 5. The scintillator single crystal according to claim 1,wherein the scintillator single crystal is expressed by (PrxLu1−x)3Al5O12 where 0.002≦x≦0.003. 6. A method of manufacturing the scintillator single crystal of claim 1, comprising charging Pr into a molten liquid expressed by (PrxLu1−x)3Al5O12 where 0.002≦x≦0.02, to an amount 5 to 15 times as much as a target amount of incorporation of Pr, and allowing said single crystal to grow by the micro-pulling-down process, using a molybdenum (Mo) crucible, or an iridium (Ir) crucible, or a crucible composed of an alloy of Ir and rhenium (Re). 7. A radiation detector having a scintillator composed of the scintillator single crystal according to claim 1, and configured as having a radiation detection unit detecting radioactive ray, and as being combined therewith a light receiving unit receiving fluorescence output as a result of detection of radioactive ray by said radiation detection unit. 8. A radiation inspection apparatus comprising the radiation detector according to claim 7. 9. The radiation inspection apparatus according to claim 8, being a positron emission tomography (PET) apparatus adoptable to a medical image processing apparatus. 10. The radiation inspection apparatus adoptable to a medical image processing apparatus according to claim 9,wherein said PET is two-dimensional PET, three-dimensional PET, time-of-flight-type (TOF-type) PET, depth-of-interaction-type (DOI-type) PET, or combinations thereof. 11. The radiation inspection apparatus according to claim 10, wherein said radiation inspection apparatus adoptable to said medical image processing apparatus is any one of stand-alone apparatus, magnetic resonance imaging apparatus (MRI), computed tomography apparatus (CT) and single photon computed tomography (SPECT), or combination thereof. 12. A radiation inspection apparatus adoptable to non-destructive inspection having the radiation detector according to claim 7, being either one of X-ray computed tomography apparatus (CT) and radiographic apparatus for radioactive ray transmission inspection, or combination thereof.
041707540	abstract	The coupling and decoupling of one or more of a plurality of differential transformers comprising a digital position probe assembly develops a phase encoded signal indicative of the position or level of a magnetically permeable element such as liquid metal or a lead screw in a nuclear reactor assembly. The phase encoded signal, which is developed as a Gray code, is processed in a phase decoder and digitally converted into a binary coded decimal to provide a digital representation of the level or position of the magnetically permeable material.
047541466	summary	CROSS REFERENCE TO RELATED APPLICATIONS The present application is a national phase application corresponding to PCT/EP 85/00669 filed Dec. 4, 1985 and based, in turn, on German Utility Models G 84 37 162.5 of Dec. 19, 1984 and G 85 28 202.2 of Oct. 3, 1985 under the International Convention. FIELD OF THE INVENTION The present invention relates to a sun-tanning apparatus comprising a bed-type supporting surface and a plurality of low-pressure A-type ultraviolet (UV-A) tubes arranged within a housing of the apparatus that are used as radiation generators, said tubes being arranged parallel to and at a distance from each other. These tubes are arranged in a trough-like reflector and are covered above by a supporting or bed surface that rests on the side walls of the housing and which is transparent to A-type ultraviolet radiation. BACKGROUND OF THE INVENTION Sun-tanning devices of this kind are known. Usually, these devices contain 10 to 20 low-pressure UV-A types, each of which is surrounded by a trough-shaped reflector, with an angle of at least 180.degree. between the axes. These reflectors are supported in the housing of the sun-tanning apparatus. The supporting or bed surface is supported on the side walls of the housing and also on the face edges of the individual reflectors that project above the low-pressure tubes. These known devices have the disadvantage that because the supporting surface lies on the face edges of the reflectors, a striped pattern results from the UV-A radiation that emerges, as almost no radiation emerges in the area of the face edges of the reflectors but only in the area of the reflector openings. SUMMARY OF THE INVENTION The present invention is directed to an improved sun-tanning apparatus of the type described above, wherein a largely even, flat radiation outlet area is provided. Thus there is provided a sun-tanning apparatus comprising a supporting surface, and a plurality of low-pressure UV-A tubes arranged within a housing of the apparatus parallel to and at a distance from each other, said tubes being located in a trough-like reflector and covered on top by a supporting surface that rests on the side walls of the housing and which permits the passage of UV-A radiation, wherein between the low-pressure UV-A tubes there are spacers distributed over the entire length and/or breadth of the apparatus in between the low pressure UV-A tubes, the supporting surface resting on the upper end of these spacers, the lower end of said spacers being secured in or on the base of the reflector. According to the present invention, spacers are arranged between the low-pressure UV-A tubes, the spacers being distributed over the length and/or breadth of the apparatus, wherein the supporting surface rests on the upper end of the spacers and the lower end of the spacers is secured in or on the base of the reflector. Thus the supporting surface is supported only in the region of the spacers so that the radiation emitted from the low-pressure UV-A tubes can pass almost completely unhindered to the outside through the supporting surface. In one embodiment the spacers are a plurality of stand-off bolts that are distributed over the length and breadth of the apparatus, the supporting surface resting on the upper end of the bolts, the lower end of the bolts being secured in the base of the reflector. The stand-off bolts are arranged in rows in the longitudinal and/or transverse direction of the apparatus. In a preferred embodiment there are four rows of stand-off bolts, located between the second, fifth, eighth and eleventh low-pressure UV-A tubes, with a total of thirteen tubes being provided. It is preferred that the stand-off bolts be in a symmetrical arrangement in order to achieve completely even support for the bed or supporting surface. However, it is also possible to use an asymmetrical arrangement, such that the stand-off bolts that make up the individual rows are staggered or are placed at irregular intervals. The stand-off bolts are in the form of studs, the lower end of which are screwed into nuts that are attached to the underside of the reflector base. A nut is provided on the inner side of the reflector and acts as a lock nut and serves to allow for precise adjustment of the stand-off distance, such that the supporting surface lies evenly on the face sides of the stand-off bolts. In a second embodiment the spacers are spacer profiles that extend wholly or in part over the length of the apparatus, the supporting surface resting on one face end of said profiles and the other end faces of the profiles being secured to the base of the reflector. It is preferred that the spacers be configured as T-spacers. The horizontal arms of the T-spacers are secured to the base of the reflector, for example by being screwed into position, by rivetting, cementing, or the like, whereas the supporting surface is installed on the face sides of the vertical arms of the T-spacers. It is advantageous that there by four spacer sections in the horizontal portion of the reflector, each of which has two or three low-pressure UV-A tubes. According to another feature of the present invention the spacer profiles are produced from material that permits the passage of UV-A radiation, so that the radiation emitted from adjacent low-pressure UV-A tubes can pass to the outside relatively unhindered. To this end it is preferred that the spacer profile be of acrylic glass. The spacer profiles are 2-4 mm thick, which is adequate to withstand the loads to which the supporting surface is subjected. It is advantageous to use low-pressure UV-A tubes that have their lower halves, as viewed in cross-section, internally silvered so that the radiation is reflected within the tubes and radiated to the outside through the supporting surface. Radiation that emerges and strikes the reflector will be reflected in the conventional manner.
	description	This present application is a divisional application of co-pending U.S. application Ser. No. 16/662,523, filed 24 Oct. 2019, which claims priority to provisional application 62/749,875, filed Oct. 24, 2018, and which is a continuation in part (CIP) of U.S. application Ser. No. 15/488,983, filed Apr. 17, 2017, which claimed priority to U.S. application Ser. No. 14/190,389, filed Feb. 26, 2014, which has issued as U.S. Pat. No. 9,636,524 on May 2, 2017, which claimed priority to U.S. application Ser. No. 13/532,447, filed on Jun. 25, 2012, now abandoned, which claimed priority to provisional U.S. patent application 61/571,406 filed Jun. 27, 2011. This invention is in the technical area of apparatus and methods for Boron Neutron capture therapy for cancer. Boron Neutron Capture Therapy (BNCT) is not new in the art, as thermal neutrons have been used for cancer therapy for the destruction of cancer tumors. These neutrons interact with boron-10 that has been placed at the cancer site. The neutrons interact with the boron to produce fission events whereby alpha particles and lithium nuclei are created. These massive ionized particles are then released, destroying the chemical bonds of nearby cancer tumor cells. At present the neutrons created in a reactor or accelerator pass through a moderator, which shapes the neutron energy spectrum suitable for BNCT treatment. While passing through the moderator and then the tissue of the patient, the neutrons are slowed by collisions and become low energy thermal neutrons. The thermal neutrons undergo reactions with the boron-10 nuclei at a cancer site, forming compound nuclei (excited boron-11), which then promptly disintegrate to lithium-7 and an alpha particle. Both the alpha particle and the lithium ion produce closely spaced ionizations in the immediate vicinity of the reaction, with a range of approximately 5-9 micrometers, or roughly the thickness of one cell diameter. The release of this energy destroys surrounding cancer cells. This technique is advantageous since the radiation damage occurs over a short range and thus normal tissues can be spared. Gadolinium can also be considered as a capture agent in neutron capture therapy (NCT) because of its very high neutron capture cross section. A number of gadolinium compounds have been used routinely as contrast agents for imaging brain tumors. The tumors have absorbed a large fraction of the gadolinium, making gadolinium an excellent capture agent for NCT. Therefore, GNTC may also be considered as a variation in embodiments of the present invention. The following definitions of neutron energy ranges, E, are used frequently by those skilled in the art of producing and using neutrons for medical, commercial and scientific applications: Fast (E>1 MeV), Epithermal (0.5 eV<E<1 Mev) and Thermal (E<0.5 eV) neutrons. BNCT has the potential to treat previously untreatable cancers such as glioblastoma multiforme (GBM). In the US brain tumors are the second most frequent cause of cancer-related deaths for males under 29 and females under 20. GBM is nearly always fatal and has, until now, no known effective treatment. There are approximately 13,000 deaths per year due to primary brain tumors. If conventional medicine is used where the glioblast is excised, new tumors almost invariably recur, frequently far from the original tumor site. Effective radiation therapy, therefore, must encompass a large volume and the radiation must be uniformly distributed. Conventional radiation treatment is usually too toxic to be of use against GBM. For distributed tumors, effective radiation therapy must encompass a larger volume and the radiation must be uniformly distributed. This is also true of liver cancers. The liver is the most common target of metastases from many primary tumors. Primary and metastatic liver cancers are usually fatal, especially after resection of multiple individual tumors. The response rate for nonresectable hepatocellular carcinoma to traditional radiation treatment or chemotherapy is also very poor. However, recent results indicate that the thermal neutron irradiation of the whole liver with a 10B compound, to be bombarded with low-energy neutrons, could be a way to destroy all the liver metastases. Recent research in BNCT has shown that neutron capture therapy can be used to treat a large number of different cancers. BNCT has been found to be effective and safe in the treatment of inoperable, locally advanced head and neck carcinomas that recur at sites that were previously irradiated with traditional gamma radiation. Thus, BNCT could be considered for a wider range of cancers. BNCT holds such promise because the dose to the cancer site can be greatly enhanced over that produced by y-radiation sources. This is a consequence of the fact that the neutron-boron reaction produces the emission of short-range (5-9 um distance) radiation, and consequently normal tissues can be spared. In addition, boron can achieve a high tumor-to-brain concentration ratio, as much as ten or more, thereby preferentially destroying abnormal tissue. BNCT has been tested using either nuclear reactors or accelerators to produce the neutrons, which are not practical or affordable for most clinical settings. Reactors also do not produce an ideal neutron spectrum and are contaminated with y-radiation. Fusion generators produce fast neutrons from the deuterium-deuterium (DD) or the deuterium-tritium (DT) reactions and are, in general, smaller and less expensive than accelerators and reactors. Fast neutrons thus produced must be moderated or slowed down to thermal or epithermal neutron energies using, for example, water or other hydrogen bearing materials. The fusion neutron generator has three basic components: an ion source, an electron shield and an acceleration structure with a target. The ions are accelerated from the ion source to usually a titanium target using a high voltage potential of between 40 kV to 200 kV, which can be easily delivered by a modern high voltage power supply. An electron shield is usually disposed between the ion source and the titanium target. This shield is voltage biased to repel electrons being generated when the positive D+ ions that strike the titanium target. This prevents these electrons from striking the ion source and damaging it due to electron heating. The target uses a deuterium D+ or tritium T+ absorbing material such as titanium, which readily absorbs the D+ or T+ ions, forming a titanium hydride. Succeeding D+ or T+ ions strike these embedded ions and fuse, resulting in DD, DT or TT reactions and releasing fast neutrons. Prior attempts at proposing fusion generators required the use of the DT reaction with the need for radioactive tritium and high acceleration powers. High yields of fast neutrons/sec were needed to achieve enough thermal neutrons for therapy in a reasonable length of time of therapy treatments. These prior schemes for achieving epithermal neutron fluxes are serial or planar in design: a single fast neutron generator is followed by a moderator, which is followed by the patient. Unfortunately, since the neutrons are entering from one side of the head, the planar neutron irradiation system leads to a high surface or skin dosage and a decreasing neutron dose deeper into the brain. The brain is not irradiated uniformly, and cancer sites have lower thermal neutron dosage the further they are from the planar port. A conventional planar neutron irradiation system 14 and its operation is shown in FIG. 1 labeled Prior Art. Conversion of fast neutrons 22 to thermal neutrons 30 takes place in a series of steps. First the fast neutrons 22 are produced by a cylindrical fast neutron generator 20 and then enter a moderating means 18 where they suffer elastic scatterings (collisions with nuclei of the moderating material's atoms). This lowers the fast neutrons to epithermal neutron 24 energies. A mixture of epithermals 24 and thermal neutrons 30 are emitted out of a planar port 16 and then enter the patient's head 26. The epithermal neutrons 24 are moderated still further in the patient's brain and moderated further to thermal neutrons, finally being captured by the boron at the tumor site. The fission reaction occurs, and alpha and Li-7 ions are released, destroying the tumor cells. The epithermal and thermal neutrons reach the patient's head through a planar port 16 formed from neutron absorbing materials that form a collimating means 28. The thermal and epithermal neutrons strike the patient's head on one side, and many neutrons escape or are not used. One escaping neutron 38 is shown as representative. This is an inefficient process requiring a large number of fast neutrons to be produced in order to produce enough thermal neutrons for reasonable therapy or treatment times (e.g. 30 min). To achieve higher yields of fast neutrons the planar neutron irradiation system 14 requires that one use either the DD fusion reaction with extremely high acceleration powers (e.g. 0.5 to 1.5 Megawatts) or the DT reaction which has an approximate 100-fold increase in neutron yield for the same acceleration power. The use of tritium has a whole host of safety and maintenance problems. Tritium gas is radioactive and extremely difficult to eliminate once it gets on to a surface. In the art of producing fast neutrons this requires that the generator be sealed and have a means for achieving a vacuum that is completely sealed. The generator head cannot be easily maintained and usually its lifetime is limited to less than 2000 hours. This reduces the possible use of this generator for clinical operation since the number of patients who could be treated would be small before the generator head would need replacement. On the other hand, the use of the DD fusion reaction allows one skilled in the art to use an actively-pumped-vacuum means with roughing and turbo pumps. The generator can then be opened for repairs and its lifetime extended. This makes the DD fusion reaction neutron generator optimum for clinical use. The downside for the DD fusion reaction is that high acceleration powers are required to achieve the desired neutron yield required by prior art methods. Improving the efficiency of producing the right thermal neutron flux at the cancer site is imperative for achieving BNCT in a clinical and hospital setting. A Boron neutron cancer treatment system is provided, comprising a moderator chamber filled with liquid or granular moderator material except for a central treatment chamber, the moderator chamber having parallel upper and lower surfaces, and a plurality of modular neutron generators fully immersed in the liquid or granular moderator material of the moderator chamber. In the following descriptions reference is made to the accompanying drawings that form a part hereof, and in which are shown by way of illustration specific embodiments in which the invention may be practiced. It is to be understood that other embodiments may be utilized, and structural changes may be made without departing from the scope of the present invention. Uniform Delivery of Thermal Neutrons to the Cancer Sites To achieve extremely high thermal neutron fluxes uniformly distributed across a patient's head, for example, a hemispherical geometry is used in one embodiment of the invention. This unique geometry arranges fast neutron sources in a circle around a moderator whose radial thickness is optimized to deliver a maximum thermal neutron flux to a patient's brain. This embodiment produces a uniform thermal neutron dose within a factor of 1/20th of the required fast neutron yield and line-voltage input power of a conventional planar neutron irradiation system. This arrangement permits using a relatively safe deuterium-deuterium (DD) fusion reaction (no radioactive tritium) and commercial high voltage power supplies operating at modest powers (50 to 100 kW). FIG. 2 is a cross sectional view of a hemispheric neutron irradiation system 36 according to one embodiment of the invention. Multiple fast neutron generators 68 surround a hemispheric moderator 34, which in turn surrounds the patient's head 26. Titanium targets 52 are distributed around the perimeter of the hemispheric moderator 34. Surrounding the moderator 34 and the fast neutron generators 68 is a fast-neutron reflector 44. In the moderator 34, moderating material such as 7LiF, high density polyethylene (HDPE), and heavy water are shaped in a hemisphere that is shaped around the head of the patient. The optimum thickness of the hemispheric moderator for irradiation purposes is dependent upon the material's nuclear structure and density. FIG. 3 shows a perspective view of a patient 58 on a table 54 with the patient's head inserted into hemispheric irradiation system 36. The patent 58 lies on the table 54 with his head inserted into hemispheric moderator 34. Surrounding the moderator is neutron reflecting material 44, such as lead or bismuth. Referring again to FIG. 2, fast neutrons 22 are produced by fast neutron generators 68. Generators 68 are composed of titanium targets 52 and ion sources 50. Ion beams are produced by ion sources 50 and accelerated toward titanium targets 52 which are embedded in hemispheric moderator 34. A DD fusion reaction occurs at the target, producing 2.5 MeV fast neutrons 22. The fast neutrons 22 enter the moderator 34 wherein they are elastically scattered by collisions with the moderator atom's nuclei. This slows them down after a few collisions to epithermal neutrons 24 energies. These epithermal neutrons 24 enter the patient's head 26 wherein they are moderated further to thermal neutron 30 energies. These thermal neutrons 30 are then captured by boron-10 nuclei at the cancer site, resulting in a fusion event and the death of proximal cancer cells. Fast neutrons 22 are emitted isotropically from titanium target 52 in all directions. Outwardly traveling fast neutrons 42 are reflected back (reflected neutron 48) by fast neutron reflector 44, while inwardly traveling fast neutrons 40 are moderated to epithermal energies and enter the patient's head 26, where further moderation of the neutrons to thermal energies occurs. A shell of protective shielding 56 is also shown in FIG. 2. In some embodiments, this may be necessary for shielding both the patient and the operator from excessive irradiation due to neutrons, x-rays and gamma radiation. The shielding can be made of a variety of materials depending upon the radiation components one wishes to suppress. In some embodiments, fast neutron reflector 44 is made of lead or bismuth. The fast neutron reflector also acts as a shielding means to reduce emitted gamma rays and neutrons from the hemispherical neutron irradiation system 36. As one skilled in the art will realize, gamma-absorbing or other neutron reflector means can be placed in layers around the hemispherical neutron irradiation system 36 to reduce spurious and dangerous radiation from reaching the patient 58 and the operator. Hemispheric moderator 34, fast neutron reflector 44 and head 26 act together to concentrate the thermal neutrons in the patient's head. The patent's head and the moderator 34 act in concert as a single moderator. With a careful selection of moderating materials and geometry, a uniform dose of thermal neutrons can be achieved across the patient's head and, if a boron drug is administered, a large and uniform therapeutic ratio can be achieved. The invention gives a uniform dose of thermal neutrons to the head while minimizing the fast neutron and gamma contributions. The required quantity of fast neutrons to initiate this performance is reduced compared to that of prior art planar neutron irradiation systems (see FIG. 1). A cross section perspective view of the hemispheric neutron irradiation system 36 in an embodiment of the invention is shown in FIG. 4. This cross-section view is of a radial cut directly through the patent's head 26 and hemispherical neutron irradiation system 36. As shown in this embodiment, ten fast-neutron generators 68 composed of ion sources 50 with titanium targets 52 are radially surrounding the hemispheric moderator 34 and the patient's head 26. The titanium target 52 in this embodiment is a continuous belt of titanium surrounding the moderator 34. The titanium targets can also be segmented, as was shown in FIG. 2. The ion sources in this embodiment are embedded in fast neutron reflector 44. There are a number of materials one could select for the moderator 34 to achieve maximum thermal neutron flux at the patient's head 26. The performance of HDPE, heavy water (D2O), graphite, 7LiF, and AlF3 was analyzed using the Monte Carlo Neutral Particle (MCNP) simulation. In general, there is an optimum thickness for each moderator material that generates the maximum thermal flux at the patient's head (or other body part or organ). The thermal neutrons/(cm2-s) was calculated for these materials as a function of moderator thickness d3, where d4=25 cm, and fast neutron reflector 44 is d1=50 cm thick and is made of lead. As in all our calculations, the combined fast neutron yield striking the area from all the fast neutron generators 68 is assumed in the MCNP to be 1011 n/s. The optimum thickness, range of thicknesses and maximum thermal neutron flux (E<0.5 eV) are given in Table I for various moderator materials. These are approximate values given to help determine the general dimensions of the moderator. TABLE IModerator ThicknessModeratorOptimumRange of thicknessMaximum FluxMaterialThickness d3 (cm)d3 (cm)(n/cm2-sec)HDPE 6 4-10 7 × 108D2O15 9-25 2 × 108Graphite2019-20 9 × 1077LiF2520-30 3 × 107AlF33020-401.5 × 107 The calculation of the therapeutic ratio is also important and depends upon the organ in question (brain, liver) and the body mass of the patient. Although HDPE gives the highest flux, it gives a lower therapeutic ratio compared to 7LiF. The designer is expected to do calculations similar to this to determine the optimum geometry for the neutron irradiation system. The MCNP simulation was used to determine the delivered dose and therapeutic ratio to the patient 58 and compare it to a planar neutron irradiation system. In one simulation, moderator 34 is composed of 7LiF whose thickness is d3=25 cm. The inner diameter of the moderator (hole for head) is d4=25 cm. The spacing between hemispheric fast neutron reflector 44 and hemispheric moderator 34 is d2=10 cm. The head is assumed to be 28 cm by 34 cm. Fast neutron reflector 44 is made of d1=20 cm thick lead in one embodiment. Thicker values of d1 increase the tumor dose rate. At a thickness of 10 cm, the tumor dose rate is about one-half the value at a thickness of 50 cm. Fast neutron generators 68 are assumed to emit a total yield of 1011 n/sec. The combined titanium targets 52 give a total neutron emission area of 1401 cm2. In the MCNP simulation BPA (Boronophenylalanine) was used as a delivery drug. The concentration of boron in the tumor was 68.3 μg/gm and in the healthy tissue was 19 μg/gm. The calculated neutron dose rates in Gy-equivalent/hr are plotted in FIG. 5 as a function of distance from the skin to the center of the head. The calculated dose rates are comparable to those used for gamma radiotherapy, typically 1.8 to 2.0 Gy per session. For the same dosage, at a rate of 3 Gy-equivalent/hr, the session length would be from 30 to 40 min. long. These session times are considered reasonable for a patient to undergo. For this simulation, the therapeutic ratio for the hemispherical neutron irradiation system is plotted in FIG. 6 as a function of distance from the skin to the center of the skull. The therapeutic ratio is defined as the delivered tumor dose divided by the maximum dose to healthy tissue. A therapeutic ratio of greater than 3 is considered adequate for cancer therapy. The conventional planar neutron irradiation system requires larger fast-neutron yields (1012 to 1013 n/s) to achieve equivalent dose rates and therapeutic ratios. In FIG. 5, a planar neutron irradiation system 14 of FIG. 1 is compared with that of a hemispheric neutron irradiation system 36 (FIGS. 2, 3, 4) in one embodiment of the present invention, using the same source of fast neutrons (1011 n/s). As can be seen from FIG. 5, the hemispherical neutron irradiation system (called radial source in FIG. 5) achieves a dose rate of about a factor of 20 over that of the conventional planar neutron irradiation system 14. The planar geometry needs a fast neutron source of 2×1012 n/s to achieve the same results. Indeed, if a DD fusion generator is used, then the planar source requires a factor of 20×increase in wall-plug power or 2.0 MW, a prohibitively large power requirement. In addition, as can be seen from FIG. 5, over a ±5 cm distance across the head center, hemispheric neutron irradiation system 36 has less than a 10% variation in dosage. A uniform dose rate is crucial for the treatment of GBM, where we want to maintain a maximum therapeutic ratio and tumors may have distributed themselves across the brain. Hemispherical neutron radiation system 36 in embodiments of the invention also gives a more uniform therapeutic ratio (FIG. 6) across the brain. The ratio is more uniform for the radial source and requires only 1/20th of the fast neutron yield of the planar source (FIG. 1). Other materials can be used for hemispheric moderator 34 in alternative embodiments. As those skilled in the art will know, high density polyethylene (HDPE), heavy water (D2O), Graphite and 7LiF can also be used. In addition, combinations of materials (e.g. 40% Al and 60% AlF3) can also be used. Different thicknesses d1 of moderator can be used to optimize the neutron flux and give the highest therapeutic ratio. The term “neutron generator or source” is intended to cover a wide range of devices for the generation of neutrons. The least expensive and most compact generator is the “fusion neutron generator” that produces neutrons by fusing isotopes of hydrogen (e.g. tritium and deuterium) by accelerating them together using modest acceleration energies. These fusion neutron generators are compact and relatively inexpensive compared to linear accelerators that can produce directed neutron beams. Other embodiments depend upon the selection of the plasma ion source that is used to generate the neutrons at the cylindrical target. These are (1) the RF-driven plasma ion source using a loop RF antenna, (2) the microwave-driven electron cyclotron resonance (ECR) plasma ion source, (3) the RF-driven spiral antenna plasma ion source, (4) the multi-cusp plasma ion source and (5) the Penning diode plasma ion source. All plasma ion sources can be used to create deuterium or tritium ions for fast neutron generation. Cylindrical Irradiation System for the Liver and Other Cancer Sites. FIGS. 7A and 7B shows another embodiment of the invention which uses a cylindrical geometry to irradiate other organs and parts of patient 58, such as the liver 76. FIG. 7A is a cross sectional view of cylindrical neutron irradiation system 62 and FIG. 7B is a perspective view of the same embodiment. In this embodiment eight fast-neutron generators 68 surround a cylindrical moderator 46. These generators 68 all emit their fast neutrons at the surface of the moderator. A cylindrical fast neutron reflector 44 surrounds the cylindrical moderator 46. As in the case of the hemispheric moderator 34, the cylindrical moderator 62 can be composed of well-known moderating materials such as 7LiF, high density polyethylene (HDPE), and heavy water. These are shaped in a cylinder that surrounds the patient. The optimum thickness of the cylinder moderator for neutron capture purposes is dependent upon the material nuclear structure and density. In this embodiment, fusion neutron generators are used to supply the fast neutrons. Fast neutron generator 68 is composed of a titanium target 52 and an ion source 50 as before. The titanium targets are contiguous to the cylindrical moderator 46. Ion beams 60 are accelerated using a DC high voltage (e.g. 100 kV) to the titanium target 52 where fast neutrons are produced from the DD fusion reaction. The fast neutrons are emitted isotropically from the titanium targets 52 on the moderator, some moving out to the fast neutron reflector 44 and others inwardly to be moderated immediately to epithermal or thermal energies. Those reflected come back into the cylindrical moderator 46 where they are moderated to epithermal and thermal energies, making their way finally to the patient 58. Cylindrical neutron irradiation system 62 permits uniform illumination of a section of the patient's body (e.g. liver) as compared to the conventional planar neutron irradiation system. In the case of the brain, the body itself acts as part of the moderation process, thermalizing epithermal neutrons coming in from cylindrical moderator 46. As one skilled in the art will realize, other cancers, such as throat and neck tumors, can be effectively irradiated by a hemispherical neutron irradiation system such as system 36. The thickness and material content of the moderator can be adjusted to maximize the desired energy of the neutrons that enter the patient. For example, for throat and neck tumors, the moderator can be made of deuterated polyethylene or heavy water (D2O) to maximize thermal neutron irradiation of the tumor near the surface of the body. For deeper penetration of the neutrons one might make the moderator out of AlF3, producing epithermal neutrons. These would be optimum for reaching the liver and producing uniform illumination of that organ. Segmented Moderator In yet another embodiment, fast neutron sources with segmented moderators may be individually moved to achieve a uniform dose across the liver or other cancer site. This geometry produces a uniform thermal neutron dose with a factor of between 1/10th and 1/20th of the required fast neutron yield and line-voltage input power of previous linear designs. This again permits the use of the relatively safe deuterium-deuterium (DD) fusion reaction (no radioactive tritium) and off-the-shelf high voltage power supplies operating at modest power (≤100 kW). A segmented neutron irradiation system 70 in an embodiment of the invention is shown in FIG. 8. Ten fast neutron generators 68, each with a wedge-shaped moderator 74, surround the patient 58. The exact shape of each moderator can vary and can be of other geometries. Each generator and moderator pair can be moved independently of the others to achieve uniformity of the neutron flux across the liver, organ, or body part. In between the wedge-shaped moderators 74 more moderating material (“filler moderating material” 72) is inserted, forming a large single moderator. The “filler” moderating material 72 can be heavy water or powered moderating materials such as AlF3. Pie shaped fillers of moderating material can also be fitted into the spaces between the wedge-shaped moderator 74. Since neutrons scatter easily, there can be some space between the wedge-shaped moderators 74 and the pie shaped fillers without undue loss of neutron moderating efficiency. The neutron yield from and the position of each fast neutron generator 68 can be adjusted to achieve uniformity across the liver or body part. The position and the neutron yield of the generator can be varied to achieve the desired radiation dose at a particular location in the patient's body. Since the cancer can be located in any part of the body, this benefit can be particularly useful for optimizing the dose at the cancer site. Surrounding the entire fast neutron/moderator system is a cylindrical fast neutron reflector 44. Fast neutrons are produced by the fast neutron generators 68 and enter the moderators 74 where they are elastically scattered by collisions with the moderator atoms' nuclei, slowing them down after a few collisions to epithermal energies. As in the other embodiments, these epithermal neutrons enter the patient 58 and liver 76, wherein they are moderated further to thermal neutron energies. The invention in various embodiments provides a uniform dose of thermal neutrons to the liver, organ or body part while minimizing fast neutron and gamma contributions. The required number of fast neutrons (e.g. 2×1011 n/s) to initiate this performance is again reduced compared to that (e.g. 2×1013 n/s) needed for the planar neutron irradiation system of the prior art. Another embodiment of the segmented design is shown in FIG. 9. The shape of the neutron irradiation system 78 is elliptical, with six sources of fast neutrons shown as distributed targets embedded in the inside elliptical moderator 96. Fast neutrons 22 are emitted isotropically in all directions. Those fast neutrons 22 moving outwardly are reflected back (see arrow 48) by fast neutron reflector 44, while fast neutrons traveling inwardly 22 are moderated to epithermal energies and enter the liver 76, where further moderation of the neutrons to thermal energies occurs. The inside elliptical moderator 96, outside elliptical moderator 98, reflector 44 and patient's body 58 act together to moderate and concentrate the thermal neutrons into the patient's liver 76. With a careful positioning of the moderators and fast neutron sources 90, 92, 94, a uniform dose can be achieved across the patient's liver, and, with a boron drug administered to the tumor, an excellent therapeutic ratio can be achieved. Elliptical neutron irradiation system 78 in FIG. 9 is a simplified cross-sectional view of the patient 58 inside the elliptical moderator 96. This cross-section view is of a radial cut directly through the patent's torso and the moderator and fast neutron generator system. To maintain visual simplicity, only the titanium targets are shown and not the ion sources. Thus, six fast-neutron sources are represented by three flat titanium targets 90, 92, 94. The rest of the fast neutron generator is not shown. Other components (e.g. plasma ion source) are neglected in the analysis. The wedge-shaped moderators 74 (used in FIG. 8) are also not shown in FIG. 9. For a simple simulation of the neutron irradiation system, the targets 90, 92, 94 are the sources of the fast neutrons and are arranged in an elliptical material 96 (e.g. AlF3, LiF). The effect of the moderating material 96, the fast neutron reflector 44 and the patient's body 58 were calculated using a Monte Carlo N-particle (MCNP5) transport code to determine how fast the neutrons were converted to thermal neutrons in the neutron irradiation system. Dosage calculations were made along a central axis of the liver. The fast neutron sources (titanium targets) are 2 cm×2 cm in area, each producing 1011/N n/s, where N is the number of sources. The human body 58 dimensions are 35.5 cm along the major axis and 22.9 cm along the minor axis. The inner elliptical moderator 96 is made of 7LiF and 10 cm thick, while the outer moderator 98 is made of AlF3 and 40 cm thick. The fast neutron reflector 44 is made of lead 50 cm thick. Boron-10 concentration is 19.0 μg/g in the healthy tissue and 68.3 μg/g in the tumor. The six sources are located in cms at: (−15, 18.06, 0) (−15, −18.06, 0) (−17, 17, 0) (−17, −17, 0) (0, 15.85, 0) (0, −15.85, 0). These measurements are made along the axis of the liver 76 from the point (−15, 0, 0) to (−5, 0, 0). In the x-direction, the first two sources 90 are centered about the left edge of the liver shown in FIG. 9, the two sources 92 are centered about the edge of the body, and the third two 94 are located above and below the origin. The origin is shown in FIG. 9 as a small cross + at the center of the body in the plane of the liver. FIG. 10 shows the therapeutic ratio for a large single dose, and the therapeutic ratio for multiple small doses (where the photon dose to healthy tissue is not included) plotted as a function of distance along the axis of the liver. The photon dose can be neglected if there is some amount of time between doses. Many of the body's healthy cells can self-repair and recover between doses. The expected therapeutic ratio is between these two curves when there is fractionation into multiple doses. In this simulation, BPA was again used as the delivery drug with the concentration of boron in the tumor at 68.3 μg/gm and in the healthy tissue at 19 μg/gm. FIG. 11 indicates that the goal of having an extremely uniform dosage to the tumor has been achieved, with about ±6% variation along the x-dimension. The calculated dose rates are comparable to those used for gamma radiotherapy, typically 1.8 to 2.0 Gy-equivalent per hour if we increase the total neutron yield to 2×1011 to 3×1011 n/s. Thus, at approximately 2×1011 to 3×1011n/s it is possible to obtain a therapeutic ratio and uniform dosage to a tumor. Approximately 10 to 20 treatments of 30 to 40 minutes would be required, with a good therapeutic ratio, uniformity of dosage, and the opportunity for healthy tissue repair between treatments. Once again, the planar neutron irradiation systems require high fast neutron yields to drive them. In one prior art system known to the inventors a fast neutron source of 3×1013 n/s is needed to obtain realistic treatment time of ˜1-2 hours. Using a D-T neutron source with a yield 1014 n/s, acceptable treatment times were obtained (30 to 72 minutes with single beam and 63 to 128 minutes with 3 beams of different direction). But these are impossible yields to achieve with realistic wall plug powers. Instead of 50 to 100 kW for the hemispheric and cylindrical neutron irradiation systems, it would take a minimum of 0.5 MW to achieve adequate yield for the planar geometry with a DT generator. These are high powers for clinics and hospitals. As one skilled in the art knows, other cancers, such as throat and neck tumors, can be effectively irradiated by the neutron irradiation system. The thickness and material content of the moderator can be adjusted to maximize the desired energy of the neutrons that enter the patient. For example, for throat and neck tumors, the moderator can be made of deuterated polyethylene or heavy water (D2O) to maximize thermal neutron irradiation of the tumor near the surface of the body. For deeper penetration of the neutrons one might make the moderator out of AlF3, producing epithermal neutrons. These would be optimum for reaching the liver and producing uniform illumination of that organ. Modular Generators As is shown in FIGS. 8 and 9, multiple modular generators may be encased in moderator material and may be arrayed to maximize thermal neutron flux at a cancer tumor location. Fast 2.5 MeV neutrons must be slowed (moderated) to energies (usually epithermal) that will penetrate to the cancer site without too many neutrons being lost in their travel to the cancer via capture by healthy tissue. These modular generators act as independent neutron sources and each may be optimized by adjustment of each individual beam's energy, direction and intensity. The modular generators can be arranged to fit a site in a particular subject's component location and structure. This is true also for cancer tumor location. The energy of the neutrons can also be adjusted by adding or subtracting moderator material. This can be done more easily than with a single beam LINAC or reactor, which usually has a fixed beamline that is integral to the neutron source. In the prior art some adjustment can be made, but the DD fusion generator in embodiments of the invention, being much smaller, can have more degrees of freedom in direction, intensity and moderation. This has an added benefit of aiding physicians in tailoring neutron radiation to the patient's cancer. Comparison to Linear Accelerators and Reactors. Modular generators in various embodiments of the present invention may also form and be part of the mechanical structure of a cancer irradiation system. This has an added benefit of moving the neutron sources as close as possible to the cancer site and the diseased body part, resulting in efficient use of the neutron source. The neutrons are being emitted in a 4π solid angle from the modular generators, so the closer to cancer site, the more of the fast neutron flux is being utilized. Linear accelerators (LINACs), which are somewhat collimated, are further from the cancer site and cannot provide this advantage. Compared to a linear accelerator, which can be several meters long or longer and may include large microwave power sources, the DD fusion sources in embodiments of the invention are less than one meter long and comprise compact microwave sources that can either be solid state microwave sources or small, inexpensive, single microwave oven magnetrons. The accelerator structure in embodiments of the invention is compact and includes a pre-moderator 118 that adds only from 5-10 cm of High-Density Polyethylene (HDPE) or 15-20 cm of polytetrafluoroethene (PTFE) Teflon to produce a first stage of neutron beam tailoring. The pre-moderator in these embodiments is an integral part of each modular generator, as is taught below with reference to several figures. In alternative embodiments other pre-moderator materials can be used such as AlF3, MgF2, 7LiF, and Fluental (trade name). Smaller, Nontoxic, Less Complex Targets for Neutron Production The modular DD fusion generator 118 in embodiments of the present invention uses a small titanium target (e.g. a 5 cm diameter disk of titanium backed by water-cooled copper fins) to produce neutrons. The target is supported directly on the pre-moderator, which is an integral part of the apparatus in this application, termed a modular generator. Linacs and other methods in the conventional arts use larger or toxic targets that require complex cooling and rotation. For example, the neutron source used by Neutron Therapeutics has a 2.6 MeV electrostatic proton accelerator and a rotating, solid lithium target for generating neutrons. In that prior-art process the Lithium becomes radioactive and toxic, and when exposed to air, it disintegrates. This prior art source has a large target chamber housing a large Li disk which is rotated in a powerful 2.8 MeV proton beam produced by a large accelerator. The Lithium wheel is roughly 2 meters in diameter and has been divided into pie-shaped sections that are removed by mechanical robotic means. In embodiments of the present invention, the Ti target is a relatively small diameter (˜5 cm) and is typically attached with 6-8 screws to the pre-moderator block and is sealed to the block with a Viton “O” ring. The Ti targets in embodiments of the invention can be easily manually removed and replaced. They also have a long lifetime and have been tested for over 4000 hours with no failures. Nuclear reactors are large structures with a substantial amount of shielding (water and concrete) and cooling systems to maintain the hot reactor core. Reactors provide primarily thermal neutrons that must be raised up in energy using an energy multiplier, and then the neutron beam must be improved to IAEA standards to produce epithermal neutrons with minimal gamma radiation. Optimizing Neutron Energy for Penetration and Minimum Damage to Healthy Tissue For tumors at depths in a subject of 3 cm or more, a goal for the moderator is to provide a neutron beam that has its energy clustered about 10 keV at the skin, in order to provide sufficient energy to penetrate a minimum of several centimeters into a human target while avoiding higher energies that are more damaging to human tissue. High conversion to epithermal energies occurs in HDPE at a thickness of approximately 5 cm, but it also produces a high yield of thermal neutrons and 2.2 MeV gammas that can damage the healthy tissue at the skin. Modular Generators In embodiments of the present invention modular generators are very important components. The modular generator combines multiple functions that were separate functions in the prior art. These integrated functions include both neutron production and beam tailoring. FIG. 12A is a perspective view of an individual modular generator 118 in an embodiment of the invention. FIG. 12B is a cross section of the modular generator 118 of FIG. 12A taken along an axis of an acceleration chamber 100 for ion beam generation and containment, and at a right angle to the axis of a turbo vacuum pump 124 that is part of the modular generator 118. FIG. 12C is a cross section of the modular generator 118 of FIG. 12A taken along the axis of the acceleration chamber 100, and along the axis of the turbo vacuum pump 124, at a right angle to the section of FIG. 12B. Each modular generator 118 can operate independently of the other modular generators and each possesses all required components to generate neutrons. Further, the various modular generators may have pre-moderators shaped to engage other building blocks of a project, such as adjacent generators or spacing moderators, as is described in enabling detail below. Viewed as in FIGS. 12A, B and C, each modular generator 118 comprises a pre-moderator 108 that is made of material known to moderate energy of energetic neutrons. In most embodiments the pre-moderator is a solid block of material, with a rather complicated shape for certain purposes. Modular generator 118 has three key elements: (1) a deuterium ion source 102, (2) an acceleration chamber 100, through which deuterium ions may be accelerated, and (3) a titanium target 106 (shown in FIGS. 12B and 12C) that is bombarded by the deuterium ions to produce high-energy neutrons. The deuterium ion source 102 has an attached microwave source 160, and microwave slug tuners 172, connected by a cable 178. Deuterium gas is leaked slowly into a plasma ion chamber 174 at the upper end of the acceleration chamber, where microwave energy ionizes the gas, creating deuterium D+ ions. The gas is ionized by microwave energy, and Deuterium (D+) ions are created and accelerated out through an ion extraction iris 138 into acceleration chamber 100, and through an electron suppression shroud 180 which deflects back-streaming electrons from being accelerated back into the plasma source, which could damage the apparatus. Electrons are being created by collisions of the D+ ions in the deuterium gas that are being created in the acceleration chamber. The deuterium ions are positively charged, and target 106 is negatively charged to a level of from 120 kV to 220 kV, and the D+ ions are strongly attracted to negatively biased target 106. Acceleration chamber 100 is connected to a turbo vacuum pump 124 that provides a modest vacuum in one embodiment of about 10−6 Torr, minimizing scattering of the D+ ions as they travel from the extraction iris 138 to the target 106. Titanium target 106 is positioned in a primary electrically insulating well 181 at the bottom of the chamber embedded into the pre-moderator material, which may be UHMW, HDPE or Teflon, of the pre-moderator 108. There is further a secondary electrical insulating well 182 surrounding the primary electrical insulating well. The surface of the moderator material in the primary and secondary electrical insulating wells may be seen as a corrugated insulator causing any surface charge to follow a curved path taken in any direction. The purpose is to provide a very long surface path to prevent electrons from traveling from the target to acceleration chamber 100 wall or any grounded element, and to avoid surface electrical breakdown or flashover in that surface path. As those skilled in the art know, the wells form an electrical insulating path. Additional corrugations or wells can be added to lengthen the path. Pre-moderator 108 has a high voltage bus bar 122 and fluid cooling channels 120 to and from the target. The high voltage is introduced via a high voltage receptacle 130 which is connected to the high voltage bus bar. Pre-moderator 108 acts as a HV insulator and as a mechanical support for the target 106 at a high negative bias. The pre-moderator 108 has metal cladding 140 at ground potential to minimize high voltage breakdown through the pre-moderator plastics. When in operation the D+ ions in the ion beam are attracted to the titanium target 106, where fast (2.5 MeV) neutrons are produced in a resulting DD fusion reaction. FIG. 13A illustrates an assembly of six modular generators 118, wherein pre-moderators 108 are spaced apart by spacers 128 which are also made of moderator material. FIG. 13B shows the arrangement of FIG. 13A in perspective. FIG. 13C shows the arrangement of FIG. 13B with one modular generator 118 removed from the assembly. FIG. 13D is a more diagrammatic illustration showing an arrangement in which modular generators may be mounted on translation and rotation mechanisms to be positioned to maximum irradiation of a cancer site. As is shown in FIGS. 13A-D the modular generators in embodiments of the invention may be arranged in an array to form a complete and moveable system of irradiating neutron sources with pre-moderators. For example, as shown in FIG. 13A-C, in the simplest configuration of the array, the modular generators may form a circle around a human torso or body part. The modular generators can be moved into three dimensional arrays around the subject to maximize neutron flux to a cancer site 148 that may not be centered on a body part 146, illustrated as a human brain in FIG. 13D. Thus, depending upon body contour, shape and size, and cancer location and distribution, the modular generators may be moved to adapt to the shape and tumor location in order to maximize the dose to the cancer and to minimize the dose to the other body parts. Referring to FIG. 13D, rotation 150 and translation 151 of the modular generator 118 can be achieved with electrical motors attached to the modular generator 118. Seven Functions of the Pre-Moderator Because the titanium target is on the pre-moderator (first stage of moderation), fast neutrons coming from the target immediately enter the pre-moderator and quickly moderated to thermal or epithermal energies. The pre-moderator also provides mechanical support, high voltage supply and cooling fluid transport to the titanium target. Exemplary pre-moderator materials that may accomplish this are Teflon and HDPE. Both Teflon and HDPE are excellent high voltage dielectrics which can also support a HV bus bar 122 and water channels 120 to be used to transport HV and the cooling fluids to the Ti target, as shown in FIG. 12C. As shown in FIGS. 12A, B, C a single generator 118 consists of an acceleration chamber 100, an ion source 102 emitting deuterium ions, a titanium target 106 and a pre-moderator 108. Pre-moderator 108 also provides a function of being a high voltage insulator for high voltage bus bar 122 that delivers high voltage (e.g. 80 kV to 300 kV)) to titanium target 106, and a water channels 120 that deliver cooling fluid to the titanium target 106. The high voltage is delivered from a high voltage power supply through a standard HV receptacle 130 to the bus bar 122 and then on to the titanium target 106, all of which are mounted in the pre-moderator 108. In various embodiments of the invention the pre-moderator 108 performs seven functions: (1) moderation, (2) mechanical support of the titanium target, (3) cooling fluid transport to the target, (4) high voltage transport to the target, (5) minimum surface flashover, (6) and a portion of a high vacuum container (a wall) with no out gassing (7). These seven attributes permit a substantial reduction of distance and amount of material between the fast neutron source and the patient, thus helping to maintain a maximum neutron flux delivered to the patient. Modular Generators Around a Subject FIGS. 13A-D show how the generators may be arranged. In FIG. 13A, six modular generators 118 form a ring around a secondary moderator 112 and are part of a structure formed by secondary moderator 112, spacers 128, and pre-moderators 108. Pre-moderators 108 and secondary moderator 112 provide the moderation function by slowing the neutrons down to epithermal energies (function #1). These elements also form a mechanical support (function #2) for the entire generator and moderator system. Secondary moderator 112 may also be a separate section attached directly to the modular generator just after the pre-moderator, each separate from the other instead of being in a ring 112 as in FIG. 13A. As shown in FIG. 12B-C, fluid transport (function #3) is supplied through channels 120, which delivers cooling fluid to target 106 to maintain the target at an acceptable operating temperature. Each generator is supplied with a separate cooling fluid input and output, wherein cooling fluid is provided through a connector 132 shown in FIGS. 12A-12C. Thus, the pre-moderator supplies fluid transport (function #4). High voltage is delivered via high voltage bus 122, which passes through pre-moderator 108 (function 4, high voltage transport). HDPE, UHMW and Teflon are excellent insulators and withstand high voltage flashover (function #6). All three may be used in vacuum systems without excessive out gassing and may help maintain the system vacuum (function #7). The achievement of these seven functions provides a very compact and flexible neutron source. The Secondary Moderator Secondary moderator 112 (FIGS. 13A-C) may comprise any one of or a combination of multiple moderator materials that optimize both the maximum flux and neutron energy for maximum dose to the cancer site. Selection (material, size and shape) may be varied depending on depth of the cancer in the subject and a desired dose at the cancer site. The secondary moderator may be D2O (heavy water) for delivery of thermal neutrons to, for example, throat and neck cancers, or a combination of AlF3 and Teflon for delivery of epithermal neutrons to brain tumors. The recommended levels of fast, thermal and gamma emission by IAEA are given in Table I. TABLE 1values in window.BNCT beam port parametersIAEA Recommended valueϕepithermal (n cm−2 s−1)~109ϕepithermal/ϕfast>20ϕepithermal/ϕthermal>100Dfast/ϕepithermal (Gy cm2)<2 × 10−13Dγ/ϕepithermal (Gy cm2)<2 × 10−13Fast energy group (ϕfast)E > 10 keVEpithermal energy group1 eV ≤ E ≤ 10Thermal energy group (ϕthermal)E < 1 eVIAEA Recommended the beam exit These IAEA recommended values depend upon older drugs, such as p-Boronophenylalanine (BPA) that have been approved for use in humans by the Food & Drug Administration (FDA) for other medical applications. Delivery of higher boron concentrations to a cancer site may depend to some extent on newer drugs to be developed, and may permit lower power, less efficient neutron beams to be used. Since treatment time might also be faster, the neutron beam quality need not be as high. DD fusion generators in embodiments of this invention have relatively low beam flux, thus permitting them to be used for cancer therapy. In some embodiments multiple modular generators may be distributed around a secondary moderator surrounding a central chamber holding a subject for treatment, providing an alternative to a completely integrated multi-ion beam system, and may have particular benefits in some circumstances. Benefits might include (1) an ability to quickly replace a single generator that has failed and needs repair; and (2) an ability to change alignment of the generators relative to one another, the moderator, and the subject. In regard to a subject, alignment of the generators may optimize dose distribution and density of neutrons at a cancer site, while at the same time minimizing spurious radiation, such as gamma rays that might be emitted external to the apparatus, or into healthy tissue of the subject. In the prior art, where reactor and accelerator neutron sources are used, careful attention has been given to achievement of high quality neutron beams to meet the IAEA standards for BNCT developed in 2001 for International Atomic Energy Agency (IAEA) (Current Status of Neutron Capture Therapy (2001) IAEA-TECDOC-1223. In embodiments of the present invention, where multiple modular DD fusion generators are used, these standards may be relaxed. The IAEA specification assumes that there is a single neutron beam that is used for all cancers and body locations. This results in standard values for the three neutron energies (thermals, epithermal and fast neutrons). Moderator and neutron spectral shifters are then designed to achieve these values for a particular fast neutron source as an input specification. This results in designs in the prior art that may not use the available fast neutrons economically and then may waste some of them to achieve the IAEA universal specs. For generators such as the DD fusion source in an embodiment of the present invention, early calculations have indicated that a single DD fusion generator would have difficulty achieving required fast neutron input to the moderating process. So, in embodiments of the invention, the use of multiple generators increases the total fast neutron yield available and allows the moderated dose to be distributed over a larger area of the body, instead of having the beam enter at one location of the body. For example, as shown in FIG. 13D, neutrons n are entering the head from many directions. This permits reduction of thermal neutron flux at any one point on the skin of the head while still achieving adequate epithermal flux to the cancer site. In early prior art reactor BNCT experiments, the thermal neutron flux burned the skin of subjects. When considering neutrons used for a particular cancer it is desirable to direct the maximum flux to the cancer site, and therefore, one must consider the specific cancer that is to be treated. This includes location and depth in the human body. Because of their relatively small size and large neutron yield, the modular generators in the embodiments of the present invention are particularly able to accomplish this by being positioned to maximize their flux at the cancer site. Since in embodiments of the invention generators are placed as close to the patient's body as practical to maximize flux at the cancer site, there is a more holistic problem. There are multiple parameters for each modular generator: (e.g. neutron flux, neutron energy, position relative to the body). What comes out of a single neutron beam pipe (1998 IAEA Standards, Table I) is not the only concern. A body part can now, in new implementations of the invention, be irradiated in all directions, and neutron intensity can be adjusted at each modular generator to achieve better flux and even more optimum neutron energy than with a single beam LINAC or a reactor. The direction of each neutron beam can be adjusted by rotating and displacing each modular generator 118. Each modular generator's yield can be adjusted electronically by varying the accelerator voltage and the ion beam current. Since the moderator size is relatively small and compact compared to the prior art, the neutron spectrum of each modular generator 118 can be adjusted by the selection of different moderator materials and thicknesses. Lowering of Required Beam Quality In embodiments of the present invention the subject's body is bombarded with neutrons from multiple directions. The neutrons can come from all sides of the body part, which minimizes the amount of distance each beam has to transverse. Unwanted neutrons striking the skin are now distributed over a larger area, reducing the skin dose of harmful components (e.g. gammas, and thermal and fast neutrons) per unit area. These components are simply delivered over a larger area of the skin. This permits adjustment of dose at the cancer site to be higher than that achieved with a single beam but with reduction of harmful components over a larger area of the skin. For a single beam case in the prior art, an argument might be made that one can rotate the patient for each exposure, but, due to possible patient movement, the neutrons would not be as accurately placed as in multi-beam embodiments of the present invention. For each placement the patient would have to be carefully re-oriented relative to the single neutron beam, which requires careful placement of the patient. In embodiments of the invention, multiple beam directions and an ability to adjust the neutron flux of each modular generator allow for optimum delivery to the cancer site while reducing harmful components. For example, if the cancer is located in the left lobe of the brain, the neutron flux to the tumor can be adjusted to deliver epithermal neutrons in the direction of that tumor. Since each modular generator neutron flux can be adjusted quickly by varying the accelerator's high voltage or the ion beam current, and by translation and rotation, this can be done easily with delivery determined by a computer program. In the present invention, a control computer monitors the ion beam current, the acceleration voltage and the output neutron yield, which can be automatically adjusted. Small modular generators in embodiments of the invention can make use of new boron drug delivery methods for higher concentrations of boron to the cancer sites. Higher concentrations of boron lower the required neutron dose and require shorter delivery time. Higher boron concentrations to the cancer site permit use of neutron generators with lower neutron yield such as the modular DD fusion generators in embodiments of the present invention. Each modular generator 118 is an independent device capable of producing neutrons independently of the other generators. This allows the total available power, P, to be distributed over N generators, resulting in the heat load being distributed safely without, for example damaging the titanium targets (unlike single target devices using lithium). In one example there are six modular generators, distributing total heat load per titanium target, since the number of neutrons per unit area is fixed by the ion beam power per unit target area. To properly treat a tumor in a subject, a large number of neutrons is required. For reasons of temperature management and stability, DD fusion generators are at present limited to fast neutron yields of less than 4×1010 n/sec. To increase the neutron yield, the number of neutron generators can be increased in embodiments of the present invention. Pre-moderators 108 can be shaped so that larger numbers of modular generators may be fitted around a subject to be treated. In the example shown by FIG. 13A there are six generators arranged equally spaced around a common secondary moderator 112, the subject cavity 116 and the subject 134. Spacing blocks 128, composed of moderator material that may be the same as that of pre-moderator 108 (e.g. Teflon or polyethylene), are placed between each pre-moderator to provide adequate spacing for fitting the subject cavity 118. The wedge angle, α, as indicated in FIG. 12A, on the pre-moderator in FIG. 13A determines the number of modules 118 with pre-moderators 108 that can fit in the circle around the patient and how close the sources may be to the patient. For example, a wedge angle of α=30° for 6 generators and α=22.5° for 8 generators. Moveable Sources with Fluid Moderator One embodiment of a system of modular generators is shown in FIGS. 13A and 13B. In FIG. 13A a plane view of six modular neutron generators 118 fitting into the cylinder (or ring) is shown. In FIG. 13B, a perspective view is shown. The modular generators can also be arranged in other patterns to maximize the dose in particular locations in the subject's body and deliver cancer therapy to selected body organs. In some embodiments of the invention the modular generators may be moved by electric motors and mechanical means to optimized locations to provide the maximum dose to the cancer site and tumor profile as determined by boron bio-distribution test biopsy and pathological analysis, Positron Emission Computed Tomography (PET-CT), Computed Tomography (CT) or magnetic resonance imaging (MRI). One may make use of moderating materials between movable modular generators. For clinical systems there should be moderator material between the modular generators. Ideally the material can quickly position itself to the new location of the modular generators and also be a moderating material. As shown in FIG. 13D, liquid moderator 156 can be used to surround the modular generators 118, acting as a secondary moderator. The moderating material is shown between the movable modular generators. The liquid is contained in an appropriate liquid container. Liquids that also have good moderating properties can be used and are easily displaced by the modular generators when moving. For example, different grades of 3M™ Fluorinert™ Electronic Liquid (e.g. FC-40), which is non-conductive, thermally and chemically stable fluid, can be inserted between generators. Like Teflon it contains primarily fluorine atoms, making it an excellent moderator, and no hydrogen. Stages of Moderation Use of multiple modular generators in embodiments of the invention permits efficient use of modulator material, reducing size of moderator and shielding material and, thus, the reduction and size of the entire system. It also reduces the required flux density of fast neutrons by bringing the neutron sources closer to the patient and directing the limited number of neutrons to the cancer site in a more efficient fashion. The subject's body also becomes part of the equation of the moderating process. The fact that the neutrons are coming from multiple directions reduces local skin dose and localized body dose of healthy tissue. Rather than coming into the body at one location, the neutrons are coming from roughly 360 degrees around the body. Moderation of fast neutrons in embodiments of the invention is a three-step process. In a first step (1) the pre-moderator 108 acts to reduce energy of the fast neutrons in as short a distance as practical with a minimal amount of gamma radiation produced in the process. The pre-moderator also serves as a medium to (2) transport high voltage and (3) cooling fluid to a fast neutron production titanium target 106. Combining these three functions ((1) moderation, (2) fluid transport and (3) high voltage transport) reduces distance and the amount of material between the fast neutron source and the patient, helping to maintain a maximum neutron flux fmally delivered to the patient. Partially slowed neutrons can then pass into the secondary moderator 112 which continues the slowing process without undue production of gamma rays from, for example, hydrogen. In the case of small animal models, the selected moderator may be heavy water (D2O). Neutron energy reduction is continued by the D2O without the generation of ˜2.2 MeV gammas that would occur if materials composed of hydrogen were used. For the case of irradiating tumors of depth greater than 3 cm in a human body, the neutrons need to be moderated to epithermal neutron energies. The human body also acts as a partial, final moderator. Thus, the epithermal energy neutrons are slowed further as they move through the body, and finally are slowed to thermal energies at the tumor site. Those skilled in the art will understand that the moderation is a statistical and random process that reduces the neutron energy with a variation or spread of the neutron energies. The process can also result in undesired gamma ray components (e.g. 2.2 MeV gammas from hydrogen capture of neutrons) which damage health cells. In embodiments of the invention, selection of the moderator material depends at least in part upon the desired energy of the neutrons at the body's skin to achieve maximum penetration to the cancer site while reducing (1) excess thermal energy components at the skin, (2) the cost and availability of the moderator material, and (3) harmful gamma ray components. Each generator's energy, yield, direction and moderation can be determined from moderation materials, the generator's voltage and acceleration current. Unlike in the prior art, dimensions of the moderator and content may be quickly changed. In some embodiments of the invention a liquid moderator (e.g. Fluorinert FC40) or a granular (e.g. AlF3) moderator may be used. The modular generators are positioned in the liquid or granular moderator material, where they are free to move by mechanical means quickly between different cancer sites. In the prior art, the moderators and shields are large, massive and usually fixed relative to a single beam reactor or linear accelerator. The patient is usually moved relative to the fixed neutron source. Using liquid or granular moderator materials permits a more efficient reduction of fast neutrons to epithermal energies while minimizing thermals and fast neutrons. Selection of the pre-moderator material is important for optimum neutron beam quality. Generally speaking, beam quality involves minimization of harmful components of radiation that accompany the production of thermal neutrons at the cancer site but also the minimization of the fast and thermal neutrons at the skin surface. In this process gamma rays are produced and, depending upon the cancer site, fast neutrons must be converted to the right energy so that they penetrate the body and deliver thermal neutrons to the tumor site with minimal irradiation of healthy tissue. Moderating the neutrons to thermal energy can result in the skin being damaged. Indeed, the thermal neutron dose to the skin can be larger than the dose to the tumor. The body itself moderates and absorbs the neutrons as they penetrate the body. Selection of the moderator material requires materials that do not moderate the fast neutrons too quickly to thermal energies. Thermal neutrons can damage the skin, and if hydrogen atoms are present in the moderation process, then damaging gamma rays are also produced. Like the moderator, the human body also moderates and absorbs the neutrons. The desired required depth of penetration depends upon the location of the tumor in the body. Simulations show that penetration of thermal neutrons starting at the skin results in penetration depths of 3 to 5 cm before a large fraction of the neutrons are absorbed. Teflon Moderator for Clinical Machine When used as a Pre-moderator, Teflon (PTFE) can satisfy 6 of the 7 functions listed above. Indeed, on several of the attributes Teflon excels. For example, since Teflon does not have atomic hydrogen, gamma production is avoided, whereas the use of HDPE does have hydrogen and, therefore maximizes the thermal neutron moderation with and added 2.2 MeV gamma ray component. Selection of HDPE as the pre-moderator material results in production of thermal neutrons in a short distance from the Ti target, whereas the use of Teflon results in a slower rate of neutron energy reduction from 2.5 MeV permitting the production of epithermal neutrons for deeper penetration into the human body and no 2.2 MeV gammas. Teflon can have a minimum high voltage in which surface arcs (flashovers or surface discharges) momentarily short out the high voltage, and lead to damage to the Teflon surface and possibly damage to the high voltage power supply. This is primarily a materials problem and not a structural problem (shape of the accelerator and Teflon shape and structure). Surface discharge along solid insulators in a vacuum in high voltage devices determines the maximum voltage between an anode and a cathode. The voltage hold-off capability of a solid insulator in vacuum is usually less than that of a vacuum gap of similar dimensions. O. Yamamoto et. al (Yamamoto, O; Takuma, T; Fukuda, M; Nagata, S; Sonoda, T “Improving withstand voltage by roughening the surface of an insulating spacer used in vacuum,” IEEE TRANSACTIONS ON DIELECTRICS AND ELECTRICAL INSULATION (2003), 10(4): 550-556) has studied a simple and reliable method to improve surface insulation strength of a dielectric such as Teflon, PMMA, and SiO2 by roughening the surface of the dielectric. Some experimental results have revealed that in a vacuum, charging along the surface of an insulating spacer precedes the flashover. The charging takes place through a process in which electrons are released from a triple junction where the cathode, insulator and vacuum meet, and propagate toward the anode, causing a secondary emission electron avalanche (SEEA) along the insulator surface. The dielectric (e.g. Teflon or HDPE) can hold charge like a battery or capacitor and then release it along the surface. This limits the use of plastics such as Teflon and HDPE as insulators and moderators inside the vacuum chamber of the neutron generator's acceleration chamber 100. For short distances across Teflon (10 mm), Yamamoto found that roughing the surface (e.g. with sandpaper or sandblasting) affects the charging of various plastics (such as Teflon and HDPE), which decreases as roughness increases. Yamamoto used roughness up to 37.8 μm but had used lower voltage gradients and smaller dielectric thicknesses (10 mm). Studies in embodiments of the present invention find that larger surfaces (distances e.g. 8 inches) of Teflon can be roughened with roughness values of 5 microns and greater and achieve high voltages of 150-220 kV for distances greater than ˜2 cm without flashover. More importantly, the roughing method gives higher insulation strengths without time-consuming conditioning previously used. This provides a significant advantage and makes generators in embodiments of the present invention operational more quickly. Depending on maximum field strength required, conditioning by the roughing process could take minutes or days. Below 1 MV m−1, the conditioning process is relatively fast. Between 1 and 10 MV m−1, the conditioning process takes longer. The best way to monitor how conditioning is going is to record the number of transient discharges (or sparks) per hour. At very high fields the arc rate might never get better than a few arcs per hour. A tolerable arc rate depends on the application. If no high voltage breakdown (arcs) can be tolerated, then the system must first be conditioned to a higher field, and then when the voltage is reduced to the operating level the arc rate drops almost to zero. For very high field strengths above 10 MV m−1, it is very difficult to condition the electrodes to give an arc rate of zero. The electrode shape and material composition becomes very important at these field levels. The Importance of the Human Body in the Moderation Process The human body acts as a moderator to reduce the epithermal neutrons to thermal energies at the cancer site. The amount of neutron energy reduction by the human body depends at least in part upon the depth of the tumor in the body. This determines the maximum neutron flux for delivery to the patient. The desired reduction of the neutron's energy will depend upon the depth of the tumor in the human body. For example, with throat and neck cancers the reduction of the neutron energy to thermal energies is desired for maximum dose to the cancer site. For small animal models, thermal energies are also desired. Dimension in the body from the skin (epidermis) to the cancer site can vary, requiring the neutron energy to be large enough for penetration to the cancer while still primarily at thermal energies, permitting capture by the boron and the destruction of cancer cells. For small animal models or skin cancer in humans, the neutrons can be at thermal energies. For cancers at deeper depths in the body, epithermal neutrons (0.025 to 0.4 eV) can be used. For deep tumors in the torso, such as, for example, pancreatic tumors, epithermal neutrons are required. Pancreatic tumors are deep in the torso and require epithermal neutrons at entrance to the body to penetrate to the tumor. Moderation of the epithermal neutrons occurs as they pass though the body. Simulations in various embodiments show that there are materials at the right thicknesses, such as Teflon, 7LiF and AlF3, which produce the epithermal neutrons that penetrate the body and thermalize by the time they reach the depth of the tumor with a maximum neutron flux. In embodiments of the invention, this occurs while minimizing production of thermal neutrons at the skin. Shape of a Clinical Machine to Match a Human Body The shape of the patient's chamber in a machine may be contoured to fit the human body part to maximize radiation to the cancer site. The shape depends upon the body part to be irradiated and the location of the tumor. As shown in FIG. 13D, for glioblastoma 148 (brain cancer), modular generators 118 may be arranged in a close ring around the head 146 that maximizes neutron flux to the cancer site 148 in the brain. The intensity of each generator can be varied to achieve maximum thermal neutrons to the tumor while minimizing the dose to healthy tissue. As discussed above, applications in embodiments of this invention permit control of the distance of each generator from the cancer site. The cancer site may be mapped using radiographic means (CT scans) and/or MRIs. A treatment planning protocol can then be determined for the optimum use of the clinical neutron source. The intensity of the neutrons coming from each neutron generator can then be varied and the location of each individual generator can be optimized. As shown in FIG. 13 D, an improvement of the moderator surrounding the modular generators is to suspend or surround the modular generators with a liquid 156 that does not contain hydrogen (a gamma producing source), but has modest atomic-number atoms like Fluorine, Carbon or Nitrogen. Various kinds of Fluorinert (tradename), FC-70 or FC-40, or FC3839 can be used. The fluid may be put between the modular generators and by mechanical means each modular generator can move independently of the other generators to a certain extent. This fluid can also absorb some heat from modular generators. As shown in FIG. 13 D, an improvement of the moderator surrounding the modular generators is to suspend or surround the modular generators with a liquid 156 that does not contain hydrogen (a gamma producing source), but has modest atomic-number atoms like Fluorine, Carbon or Nitrogen. Various kinds of Fluorinert (tradename), FC-70 or FC-40, or FC3839 can be used. The fluid may be put between the modular generators and by mechanical means each modular generator can move independently of the other generators to a certain extent. This fluid can also absorb some heat from modular generators. Generator Alignment In embodiments of the present invention each stand-alone generator, as seen in FIG. 13D, for example, may be positioned and aligned to give a maximum flux and neutron distribution at the cancer site. Each generator is small enough in size and weight that the generators may be mechanically moved and positioned so that optimum neutron flux at the cancer site is achieved, depending upon the cancer's location and distribution. The generators may be arranged around a moderator whose radial thickness is optimized to deliver a maximum thermal neutron flux to the cancer site. Depending upon the body part being irradiated, the geometry can be circular or elliptical. By selecting the moderating material and radial thickness one can deliver thermal neutrons to the cancer site. FIG. 14A shows an on-axis view of an exemplary clinical neutron source using multiple modular generators 118 for BNCT of a human head. This example uses eight modular generators 118 and assorted moderator materials coupled with reflecting and shielding material (e.g. graphite 144). Secondary moderators (166 and 170) can be composed of one or more materials. There are moderator spacing blocks 128 in one embodiment composed of the same material High Density Polyethylene (HDPE), Ultra High Molecular Weight polyethylene (UHMW), or (PTFE (Teflon)) as the secondary moderators. Blocks of these materials fit in between the modular generators and are adjacent to each generator's pre-moderator. They act as mechanical spacers as well as moderator components. The outside of this region, between and behind the modular generators 118, is filled with either graphite or lead 144 to serve as a neutron reflector and shield. FIG. 14B also shows a side section view of the apparatus of FIG. 14A taken along a line through the top and bottom generators. There is additional moderator material in the front and behind the modular generators, extending a little above the pre-moderator. In our example, the cylindrical space 164 available for the patient's head is 52 cm deep and 30 cm in diameter. This space might be lined with 1-mm of cadmium 162 to shield against too large a thermal neutron dose to the patent's skin. Shield 162 is also shown in FIG. 14A. In other embodiments this space may be lined with 6LiF. The exemplary arrangement as illustrated in FIGS. 14A and B has a secondary moderator consisting of multiple layers of 40% Al and 60% AlF3 (166) and an additional moderating cylinder 170 of either 7LiF or D2O. These materials are shown to be concentric rings in FIG. 14A. Since 7LiF or D2O can be expensive, thicknesses were varied to obtain a desired neutron beam quality without over-using either 7LIF or D2O. In the example shown in FIGS. 14A and 14 B the thickness ratio between the two segments is altered, the total moderator thickness is 34 cm, and the sources are R=52.5 cm from the origin (center of the brain). The effect of doing this varying these materials is plotted graphically in FIG. 15. The reflector material graphite 144 is 30 cm thick in this example, the thickness of the Teflon 168, t, in front of the 2.5 MeV source is varied, and the portion 170 of the moderator is either 7LiF or D2O. As t changes, the thickness of the Al/AlF3 166 of the moderator changes, with all other dimensions remaining constant. The target is embedded in the Teflon 168, UHMW or HDPE. Sources are titanium targets 106 being bombarded by deuterium ion beams 5.0 cm in diameter. Each target is emitting 4×1010 neutron/sec. Eight modulator generators 118 emit 3.2×1011 n/s total emission. A concentration of 10B in the tumor and health tissue (e. g. skin) is known to be possible. 10B tumor concentration is assumed to be 50 ppm, while 10B in healthy tissue is 15 ppm. The relative biological effectiveness (RBE) for 10B in tumor is 2.7, and in healthy tissue is 1.3. Tumor and healthy tissue doses are calculated using the NRC and ICRP models for neutron RBE. The material 7LiF was the best performer and D2O was second best. An important main objective in these examples is to give a sufficient dose of neutrons to the cancer while minimizing the dose to the healthy tissue and not damaging it. FIG. 15 shows the performance for moderators with different values for tin cm and either 7LiF or D2O in the secondary moderator. The ordinate R is the ratio of tumor dose at the origin to healthy tissue skin dose, and the tumor dose at the center of the brain assumed to be the site of the cancer. As can be seen from FIG. 15, 7LiF outperforms D2O. The best performance is R=1.9 and a tumor dose in excess of 1.4 Sv/hr. A consequence of RBE is that a small percentage of fast neutrons is essential to obtain a high value for R; also, a reasonable number of epithermals is required to penetrate the target. Thus a combination of 7LiF and D2O may outperform either material alone. A Need for Small Animal Neutron Sources Development of boron delivery agents for BNCT is an ongoing and challenging task of high priority. A number of boron-10 containing delivery agents have been prepared for potential use in BNCT. With the development of new chemical synthetic techniques and increased knowledge of the biochemical requirements needed for an effective agent and their modes of delivery, a wide variety of new boron agents has emerged, but only two of these, oronophenylalanine (BPA) and sodium borocaptate (BSH) have been used clinically and have US FDA approval. Patient-derived xenograft (PDX) is created by transferring primary tumors from a patient into a mouse or small animal model. Tests of delivery and effectiveness of drugs to the cancer site can then be performed. In the prior art, only beamlines from nuclear reactors and linear accelerator structures can be used. A small laboratory neutron source, as in embodiments of this invention, is therefore valuable in the development and testing of new boron delivery drugs and their effectiveness in destroying the cancer site. As compared to a clinical delivery system, a smaller number of stand-alone generators such as generators 118 is needed for a delivery system for a small animal such as a mouse. The modular generators used have a slab wall angle of α=0 (see α defined in FIG. 12 A). The secondary moderator may be a separate container of heavy water (D2O). Since the small animal target is indeed small, the secondary moderator volume can be reduced, and the compact modular generators can be moved close to it permitting the modular generators to be closer to the animal target. Thus, the neutron flux at the cancer site is increased, and with proper selection of moderator material and size, will still be able to moderate the neutrons to IAEA standards. In addition, by moving closer, the number of generators can be reduced while still maintaining a high thermal neutron flux at the cancer site. In our example of the new art for a small animal source, we can use four modular generators 118 to emit enough thermal neutrons at the cancer site. We can use the modular generators of 12 A, B, C with the slab wall angle of α=0. This makes the pre-moderator 108 a rectangular cuboid (or “rectangular slab” of). FIG. 16A is a perspective view of a modular generator having such a rectangular pre-moderator 108, making it suitable for arrangements of four generators in a rectangular array, as shown in FIG. 16B. In FIG. 16B the four modular generators are arranged around a secondary moderator 112, which in one embodiment may be a container of heavy water or granulated moderator material. FIG. 16C is a cross section view of the arrangement of FIG. 16B, taken along section line4 16C-16C of FIG. 16B. The elements previously annotated for modular generators are reused in FIGS. 16 A, B and C. FIG. 16D is an exploded view where the four generators 118 are moved back from the small heavy water moderator 112. Each generator 118 has a pre-moderator 108 with a fast neutron generator with a titanium target 106. A deuterium ion beam is generated by a plasma ion source 102 and accelerated in an acceleration chamber 100 to the titanium target 106, where the DD fusion reaction occurs releasing fast 2.5 MeV neutrons. This description is all common to the descriptions or other embodiments in the specification. The neutrons generated pass through a pre-moderator 108, where they are partially moderated to thermal neutron energies. They then pass into the moderator block 112 where they are further moderated, reducing the energy of fast neutrons to thermal neutron energies. The thermal neutrons then enter a cylindrical mouse chamber 114 where they enter the small animal 116. The pre-moderator is designed to slow the fast neutrons to thermal neutrons by scattering the fast neutrons via collisions with the hydrogen in the HDPE or UHMW plastics. The distance the 2.5 MeV neutrons have to traverse is approximately 3 to 5 cm, wherein approximately 50% of the neutrons lose enough of their energy to be classified as thermal neutrons. These neutrons, containing both thermal and fast neutron components, can then travel into the moderator box 112, where they are further moderated by collisions with deuterium atoms. Roughly speaking, the HDPE with its hydrogen-atoms moderates the neutrons to thermal energies over a short distance; the thermalized neutrons then penetrate the cylindrical chamber 114 wherein they place the small animal 116. The small animal model is used to test the delivery of boron to the cancer site. For the pre-moderator, high density polyethylene (HDPE) is optimum for producing the maximum flux of thermal neutrons. As in the case of the clinical generator, it is desired to produce a maximum thermal flux at the cancer site. A mouse is a small object, and penetration of thermal neutrons to the cancer site can easily be achieved. Moderation of the fast neutrons to thermal energies is desired with minimum production of gamma radiation, which is harmful to the healthy cells. As those skilled in the art will understand, hydrogen atoms are excellent at scattering fast neutrons, resulting in moderation of the neutrons to thermal energies in the shortest path length in the moderating material. Indeed, using 5-6 cm of high-density polyethylene (HDPE) or UHMW plastic results in moderation of about 50% of 2.5 MeV neutrons to thermal energies. Further moderation of the neutrons by longer distances in the HDPE results in more fast neutrons being converted to thermal energies. However, this results in reduction of the total flux (n/cm2) that is available since the neutrons are being emitted in a 4π solid angle. Hydrogen capture of neutrons produces high energy gamma radiation, which is destructive to both healthy and cancerous cells. Adding another moderator to further thermalize the neutrons is accomplished by the use of heavy water (D2O). The skilled person will understand that the embodiments described in this application are exemplary, and not limiting. Many variations may well fall within the scope of the invention, which is limited only by the scope of the following claims.
047740493	description	DESCRIPTION OF THE PREFERRED EMBODIMENT FIG. 1 shows a schematic representation of the nuclear steam supply system 100 of a typical pressurized water reactor (PWR) 1 for generating electric power which can employ the method and apparatus of the present invention to more accurately monitor the radial and three dimensional power profiles within the core to avoid the operating difficulties experienced by the prior art. As shown in FIG. 1, PWR 1 includes a vessel 10 which forms a pressurized container when sealed by its head assembly 11. The vessel 10 has coolant flow inlet means 16 and coolant flow outlet means 14 formed integral with and through its cylindrical walls. As is known in the art, the vessel 10 contains a nuclear reactor core 5 of the type previously described consisting mainly of a plurality of clad nuclear fuel elements 20 (only a few of which are shown) arranged in assemblies, for example assemblies 22 and 24, which generate substantial amounts of heat, depending primarily upon the position of the full length control rods such as 14. Fission reactions within the core 5 generate heat which is absorbed by a reactor coolant, for example light water, which is passed through the core 5. The heat is conveyed from the core by coolant flow entering through inlet means 16 and exiting through outlet means 14. Generally, the flow exiting through outlet means 14 is conveyed through an outlet conduit, hot leg 9, to a heat exchange steam generator system 26, wherein the heated coolant flow is conveyed through tubes which are in heat exchange relationship with water which is evaporated to produce steam. The steam produced by the generator is commonly utilized to drive a turbine-generator 42 to produce electric power. The cooled reactor coolant is conveyed from the steam generator 26 through a cold leg conduit 15 by reactor coolant pump 32 to inlet means 16. After being delivered to the reactor pressure vessel 10 through inlet means 16, the coolant is forced to circulate downwardly around the outside of a core barrel assembly 12 and upward through the interior of the core 5, through the coolant channels formed by the assemblies, whereby the reactor coolant cools the core 5 and its fuel rods 20. A pressurizing system (not shown) is provided to maintain the pressure of the reactor coolant with certain acceptable limits. Thus, a closed recycling primary loop is provided with the coolant piping coupling the vessel and the steam generator. The vessel illustrated in FIG. 1 only shows one steam generator for clarity, however, it is adaptable for more than one such closed fluid system or loop, though, it should be understood that the number of such loops varies from plant to plant and commonly two, three, or four are employed. The reactivity of reactor core 5 is controlled by dissolving a neutron absorber, such as boron, in the reactor coolant, and by the insertion into the core of control rods, for example control rod 14. The boron concentration of the reactor coolant is regulated by a reactor makeup system 19 which extracts coolant from the cold leg 15 upstream of the reactor coolant pump 32, adds or removes boron as appropriate, and returns the coolant with the proper boron concentration to the cold leg 15 downstream of the pump 32. The control rods, such as control rod 14, which are made of neutron absorbing material, are inserted into and withdrawn from the reactor core 5 by a rod control system 21. The rod control system 21 receives commands from the reactor control system 23. Typically, control rods are moved in groups, which groups are referred to as control rod banks. This rod control system 21 is well known in the art and provides one measure of the position of every control rod. Ex-core neutron detectors such as detectors in detector systems 35 and 36 monitor the neutron flux, and therefore the power level of reactor core 5. In addition, most PWRs are provided with an in-core movable detector system 27 for detecting neutrons which includes a number of thimbles 29 distributed across the core 5 through which movable detectors 31 may be inserted to generate a detailed map of the power distribution in the core. Such mapping is performed periodically, such as monthly, to determine if there are any potential or actual limiting hot spots in the core. Some PWRs are provided with a fixed in-core detector system (not shown) in which strings of detectors are permanently positioned in thimbles similar to thimbles 29. If the number of strings of fixed in-core detectors installed in the reactor is sufficiently large (on the order of forty or more) these installations do not require the in-core movable detector system 27. However, the costs associated with the installation and use of such many-string fixed in-core detector systems tend to be relatively large and so these systems are not universally used. If the number of strings of fixed in-core detectors installed in the reactor is sufficiently small (not exceeding eight in typical installations), the associated costs are modest, but the availability of an in-core movable detector system appears to be unavoidable. FIG. 2 shows core map 197 which is a plan view of the reactor core 5 comprising fuel rod assemblies 62. In present practice, there are thermocouples installed at or just above the outlet nozzles of a fraction of the fuel assemblies in most commercial pressurized water nuclear power reactors. These thermocouples will be referred to hereinafter as core-exit thermocouples. Typical reactor cores generally consist of from approximately one hundred to more than two hundred assemblies and the thermocouples are usually located at approximately one out of four fuel rod assemblies. FIG. 3 shows a plan view of reactor core 5 which illustrates the relative position of the fuel rod assemblies 62, along with an exemplary arrangement of the core-exit thermocouples 64 within the fuel assembly locations 62. The outputs from the core-exit thermocouples 64 are sent to core-exit thermocouple system 71. Typically, the core-exit thermocouple system 71 periodically samples the thermocouple voltages and converts them to digital values in convenient engineering units, .degree.F. or .degree.C. Temperatures corresponding to the electrical outputs from core-exit thermocouples 64 are presently printed on a line printer by the supervisory plant computer (not shown). However, information in this form does not lend itself to easy interpretation by the plant operator and is not otherwise used. For more effective information transfer, the first aspect of present invention provides a coherent pictorial presentation of the radial power distribution for the fuel rod assemblies 62. The visual display employed utilizes a color display over a core map, similar to the core map 197 shown in FIG. 2. In one embodiment of the present invention, the display comprises deviations of the core power from reference values provided by the core flux map which is periodically made or from reference values from a previous thermocouple-generated power distribution. If there is no deviation, within certain preset limits, then the color chosen to display that fact is green. As the assembly power decreases from that of the equivalent reference, the color changes to add more and more blue. Conversely, when the assembly power increases from that of the reference, the color changes to add more and more red so as to change from yellow to orange to red. The active display area is divided into blocks corresponding to the relative location of the fuel assemblies 62 within the core 5 or to subregions, such as quarters of fuel assemblies at the respective relative locations. As described above, in present day reactors, there are approximately one-fourth the number of core-exit thermocouples as fuel rod assemblies. In order to have a coherent display, hypothetical temperature signals must be derived for the other three quarters of the fuel assemblies. For this purpose, a surface spline algorithm is used to interpolate for assemblies or subregions of assemblies not directly covered by thermocouples and thereby to provide a coherent display. Additionally, visual and/or audio alarm provisions are included to indicate when either deviations from reference values or inferred absolute values of power distribution fall outside preset limits. In accordance with the above, a first aspect of this invention may be generally understood by reference to the block diagram of FIG. 4. Each of the individual blocks illustrated in FIG. 4 is set out in greater detail in the figures following hereafter. In addition it will be apparent that the preferred embodiment of this aspect of the present invention comprises a set of computational algorithms and computational control logic embodied in computer software which is executed on a digital computer. Block 71 represents a core-exit thermocouple system which receives the outputs of the thermocouples, as shown in FIG. 3. The signals derived from the thermocouples, representative of the relative temperatures encountered within the corresponding core locations, are fed from Core-Exit Thermocouple System 71 to a Signal Processing Unit, represented by block 72 (As described below, Signal Processing Unit 72 comprises Thermocouple F.sub..DELTA.H Calculator 91 and F.sub..DELTA.H Deviation Calculator 92). Signal Processing Unit 72 processes the signals into compatible form to interface with the Interpolator 73 (As described below, Interpolator 73 comprises F.sub..DELTA.H Deviation Interpolator/Extrapolator 93, Current Assemblywise F.sub..DELTA.H Synthesizer 106, Power Distribution Deviation Calculator 103, and Relative F.sub..DELTA.H Deviation Classifier 94). The respective signals, representative of the power of the assemblies, are then fed to Display Interface 74 (As described below, Display Interface 74 comprises Two-Dimensional Graphics Generator 95) which processes the signals into a form compatible with the Display Device 75 (As described below, Display Device 75 comprises Graphics Monitor 100). As shown in FIG. 5, calibration factors for use in embodiments of the present invention are calculated on a periodic basis in Core-Exit Thermocouple Calibrator Module 80 by inputting thereto measurements from (1) In-Core Movable Detector System 27 in response to data received from movable detectors 31 or from Fixed In-Core Detector System, if provided; (2) Core-Exit Thermocouple System 71 in response to data received from core-exit thermocouples 64; (3) Reactor Control System 23 in response to data received from hot-leg temperature detector 37 and cold-leg temperature detector 38; and (4) Reactor Protection System 41 in response to data received from pressure system 51-53. In accordance with methods well known in the art, Core-Exit Thermocouple Calibrator Module 80 computes and stores the following information for use in the manner to be explained in detail below: (1) a library 84 of self consistent calibration factors for the individual core-exit thermocouples 64; (2) a library 85 of reference "assemblywise" relative enthalpy rise values, F.sub..DELTA.H, for all fuel assembly locations in the core, which reference values are obtained from a core flux map provided by In-Core Movable Detector System 27 or, if provided, by a Fixed In-Core Detector System; and (3) a library 86 of reference relative enthalpy rise values, F.sub..DELTA.H, for the respective functioning core-exit thermocouples 64 in the reactor vessel, which reference values are obtained from a core flux map provided by In-Core Movable Detector System 27 or, if provided, by a Fixed In-Core Detector System. The contents of library 86 are in fact a subset of the contents of library 85. These data in libraries 84 through 86 and in libraries 87 through 88, noted later, are stored for later use. In one embodiment of the present invention the libraries are stored in a computer memory, in another embodiment, they are stored on computer peripheral equipment such as on magnetic disk storage. Libraries 84 through 86 are updated infrequently, typically once each effective full power month of plant operation. The apparatus and methodology of systems 23, 27, 38, 41, 51-53, 71, and 80 used to create libraries 84 through 86 are well known in the art, and, as such, are not included in this disclosure. The details of the operation of this aspect of the present invention are now described with reference to FIG. 6. Core-Exit Thermocouple System 71 generates the actual coolant core-exit temperature values used in the inventive system in real time, i.e., as the measurements are received from the core-exit thermocouples 64; the core-exit thermocouples 64 are passive and operate continuously. Core-Exit Thermocouple System 71, which processes the output signals from the core-exit thermocouples 64, operates frequently, and in real time, in response to (1) a preset elapsed time signal, typically on a 30 second update cycle, or (2) a preset change in control bank position or (3) a change in power level signal, typically, 10 steps of control bank movement or 3 percent in power level, respectively, and initiates operation of the remainder of the system. Again the methodology and apparatus of Core-Exit Thermocouple System 71 is well known in the art and, as such, is not included in this disclosure. The core-exit temperature values computed in Core-Exit Thermocouple System 71; the hot leg temperature and cold leg temperature obtained from resistance temperature detectors 37 and 38 shown in FIG. 1, respectively, and passed through Reactor Control System 23; and reactor coolant system pressure obtained from pressure system 51-53 shown in FIG. 1 and passed through Reactor Protection System 41; are all input into Thermocouple F.sub..DELTA.H Calculator 91. Thermocouple F.sub..DELTA.H Calculator 91 then converts the inputs into values of the difference between the enthalpy of the coolant leaving the core and that of the coolant entering the core at the fuel assemblies at which actual thermocouples are positioned, using the library of thermocouple calibration factors 84 and methods well known in the art. The F.sub..DELTA.H values generated in Thermocouple F.sub..DELTA.H Calculator 91, F.sub..DELTA.H.sup.j(obs), and reference F.sub..DELTA.H values from library 86, F.sub..DELTA.H.sup.j(ref), are transmitted to F.sub..DELTA.H Deviation Calculator 92. F.sub..DELTA.H Deviation Calculator 92 then computes the fractional deviation of the currently observed F.sub..DELTA.H values for each functioning core-exit thermocouple 64 from the corresponding reference value according to the relation: EQU dF.sub..DELTA.Hj =(F.sub..DELTA.H.sup.j(obs) -F.sub..DELTA.H.sup.j(ref))/F.sub..DELTA.H.sup.j(ref) (1) where j signifies any given functioning core-exit thermocouple. The fractional deviation values for each functioning thermocouple 64 are passed to the F.sub..DELTA.H Deviation Interpolator/Extrapolator 93 which uses the fractional deviation values, together with stored values of the geometric coordinates of the locations of the respective core-exit thermocouples, to generate a mathematical function that characterizes the distribution of the fractional deviations of F.sub..DELTA.H over the entire reactor cross section, and subsequently provides the estimated F.sub..DELTA.H fractional deviation value at each fuel assembly location and/or at the center of each quarter of each fuel assembly in the core. Although many different mathematical techniques could be used to provide the estimates, a preferred mathematical technique used to generate the estimates is referred to in the literature as the "Surface Splines Method" and is in common use in a variety of two dimensional interpolation applications. The deviation values of the observed parameters, as well as the estimated deviations, are transmitted to the Current Assemblywise F.sub..DELTA.H Synthesizer 106. The Current Assemblywise F.sub..DELTA.H Synthesizer 106 determines the assemblywise values of power, F.sub..DELTA.H, using the Library of Reference Assemblywise F.sub..DELTA.H Values 85. These assemblywise values power, F.sub..DELTA.H, are then provided to Power Distribution Deviation Calculator 103. The Power Distribution Deviation Calculator 103 generates the deviation values, dF.sup.XY.sub..DELTA.H, for display using the Library of Display Reference Power Distribution 89. On command, Reference Updater 105 replaces the stored values in Library of Reference Display Power Distribution 89 with the current thermocouple based power distribution produced by the Current Assemblywise F.sub..DELTA.H Synthesizer 106 or the Library of Reference Assemblywise F.sub..DELTA.H Distribution 85. By periodically exercising Reference Updater 105, either upon manual initiation or automatically on periodic update, during periods of reactor maneuvering, the operator can avoid generating indications of what would otherwise appear to be significant anomalies in power distribution when in fact, the so-called anomalies are simply the result of conventional operations under the operator's control, such as control bank insertion during a load reduction. The deviation values, dF.sup.XY.sub..DELTA.H, produced by Power Distribution Deviation Calculator 103 are then transmitted to Relative F.sub..DELTA.H Deviation Classifier 94. The Relative F.sub..DELTA.H Deviation Classifier 94 classifies the deviation value at each fuel assembly or quarter of a fuel assembly location according to the sign and magnitude of the deviation value and a class identification parameter is associated with the corresponding location indices. In the preferred embodiment, a total of 11 classifications are considered. The spectrum of fractional deviation values covered by the set of classes is continuous, with the upper bound of one class being also the lower bound of the next higher class. The lowest and highest classes, class 1 and class 11, in this embodiment, are each unbounded on one side so as to encompass all "extreme" deviation values. Two-Dimensional Graphics Display Generator 95 relates the class identification parameter value to a specified color, using, in this embodiment, a look-up table. Two-Dimensional Graphics Display Generator 95 generates a block of the appropriate color corresponding to a fuel assembly or quarter of a fuel assembly at the appropriate relative location on a graphics monitor 100. In this embodiment green was chosen to correspond to class 6 with boundaries of approximately .+-.0.03 in F.sub..DELTA.H fractional deviation. Thus, if no anomalies in core power distribution exist, i.e. the current core wide F.sub..DELTA.H distribution is very similar to the reference F.sub..DELTA.H distribution, the entire core cross section on the display screen would be a uniform green color. FIG. 7 shows a display corresponding to the case where a control rod located at coordinates D-12 was partially inserted into the core. This insertion depresses power generation in the vicinity of the control rod and causes a broadly based, moderate increase in power generation on the opposite side of the core since in this case total power output was maintained constant despite the insertion of the single control rod. For human factors purposes, progressively larger negative deviations are assigned progressively darker shades of blue; progressively larger positive deviations yield progressively brighter colors, passing from yellow through orange to red. Thus, the transition from an all green core cross section display to one containing regions or contours of other colors provides an immediate visual indication to the reactor operator that the core radial power distribution has changed. The specific colors that appear are indicative of the relative severity of the change and the distribution of the regions and/or contours provides a readily recognized "signature" of the type and probable cause of the change. In the illustration shown in FIG. 7, region 501 is light blue, region 502 is made up of progressively smaller nominally concentric areas of progressively darker blue, region 503 is yellow, and region 504 is light orange. In an additional feature of this aspect of the present invention, the array of class identification parameter values generated by Relative F.sub..DELTA.H Deviation Classifier 94 is periodically and automatically scanned by Alarm Assessor 96, and an audible and/or visual alarm is generated by Control Room Alarm 97 upon detection of a class identification parameter lying outside a predetermined acceptable range. The alarm draws the operator's attention to the fact that an anomaly of some sort exists in the core power distribution and that displays of the current deviation values and of the current absolute values should be viewed for diagnosis and evaluation. In yet a further additional feature of the present invention, the core wide estimated F.sub..DELTA.H fractional deviation distribution values generated by F.sub..DELTA.H Deviation Interpolator/Extrapolator 93 are combined with the values of F.sub..DELTA.H in Reference Assemblywise F.sub..DELTA.H Distribution Library 85 in Current Assemblywise F.sub..DELTA.H Synthesizer 106 to generate current estimates of the distribution of absolute F.sub..DELTA.H values. An appropriate mathematical expression describing this process is: EQU F.sup.current .sub..DELTA.H;k,1 =F.sup.ref.sub..DELTA.H;k,1 * (1+.sigma.F.sub..DELTA.H;k;1) (2) where the indices k and 1 identify the geometric location in the core cross section at which the value of F.sup.current.sub..DELTA.H;k,1 exists. The array of current F.sub..DELTA.H values and the appropriate location data are passed to Absolute F.sub..DELTA.H Classifier 107. Absolute F.sub..DELTA.H Classifier 107 sorts the absolute F.sub..DELTA.H values into classes in the same manner as was described above with respect to Relative F.sub..DELTA.H Deviation Classifier 94. The array of absolute class identification numbers is then passed to the Two Dimensional Graphics Display Generator 95 and is converted into a multicolor core cross-section display in which the "hotter" assemblies are indicated by progressively brighter colors, with red indicating "very hot" assemblies which are characterized by F.sub..DELTA.H values greater than a preset limit and the "cooler" assemblies, by progressively darker shades of blue. With this display an operator can immediately assess whether the core contains any assemblies that are operating at a relative power level greater than that permitted in the plant Technical Specifications. In yet a still further additional feature of this aspect of the present invention, the array of class identification parameter values generated by Absolute F.sub..DELTA.H Classifier 107 is periodically and automatically scanned by Alarm Assessor 96, and an audible and/or visual alarm is generated upon detection of a class identification parameter lying outside a predetermined acceptable range. The alarm draws the operator's attention to the fact that an anomaly of some sort exists in the core power distribution and that displays of the current deviation values and of the current absolute values should be viewed for diagnosis and evaluation. A second aspect of this invention may be generally understood by reference to the block diagram of FIG. 8. Each of the individual blocks illustrated in FIG. 8 is set out in greater detail in FIG. 9. Again, it will be apparent that the preferred embodiment of this second aspect of the present invention comprises a set of computational algorithms and computational control logic embodied in computer software which is executed on a digital computer. Referring to FIG. 8, block 200 represents a conventional ex-core neutron detector system consisting of one or more sets of two or more neutron sensitive nuclear detectors, for example detector sets 35 and 36 shown in FIG. 1, wherein the detectors, sometimes referred to as the "sections," of each set are arrayed one above the other over a span roughly comparable to the height of the core. Alternatively, block 201 represents a system comprised of a few (typically, four or eight) strings of fixed in-core neutron or gamma ray sensitive nuclear detectors that are arrayed in a specified pattern, either among or within selected fuel assemblies in the core. Each string consists of several (typically, four to seven) individual sensors mounted one above the other in a common thimble or tube. Either the Ex-Core Neutron Detector System 200 or the Few-String Fixed In-Core System 201 is required; however, both systems would not commonly be used concurrently. The signals derived from the sensors in block 200 or block 201, representative of components of the axial nuclear power distribution in the region or regions of the core monitored by the respective detector sets or strings, are directed to Nuclear Detector Signal Processing Unit 202 (As described below, the Nuclear Detector Signal Processing Unit 202 comprises Axial Power Distribution Synthesizer 301, Radial Power Distribution Expander 302, and Power Distribution F.sub..DELTA. H Calculator 303). The Nuclear Detector Signal Processing Unit 202 processes the input ex-core or incore detector signals into compatible form to be passed to Power Distribution Modifying Unit 203 (As described below, the Power Distribution Modifying Unit 203 comprises F.sub..DELTA.H Deviation Calculator 304, F.sub..DELTA.H Deviation Interpolator/Extrapolator 93 and Power Distribution Adjuster 305). As shown in FIG. 8, Core-Exit Thermocouple System 71 and Thermocouple F.sub..DELTA.H Calculator 91 are identical to, and perform the same functions as, the identically numbered components described above. The output signals from the Power Distribution Modifying Unit 203, which are representative of the synthesized current three dimensional power distribution in the reactor core, are passed to Display Interface 204 which processes the signals into a form compatible with Display Device 205. This second aspect of the present invention utilizes the Core-Exit Thermocouple Calibrator 80, described above, in the same manner as noted above, to extract a library of Thermocouple Calibration Factors 84 from the results of sets of concurrently measured core flux maps and core-exit temperature maps. In addition, a Nuclear Detector Calibrator (not shown) generates a Library of Axial Expansion Coefficients 87. The methodology and techniques used to determine the values of the expansion coefficients depend to some degree on the particular type of nuclear detector used, but all utilize information derived from flux maps as input. All relevant methodologies and techniques are well known in the art. A Library of Radial Power Ratios 88 is generated by application of simple editing methods to the results either of one or more flux maps obtained in the applicable reactor core or of detailed three dimensional calculations using an analytical model representative of the applicable reactor core. The methodology for obtaining such a library is well known to those of ordinary skill in the art. The details of the operation of this second aspect of the present invention are now described with reference to FIG. g. Ex-core Neutron Detector System 200 or Few-String Fixed In-Core Detector System 201 generates nuclear detector signal values in real time, i.e. as signals are received from sensors such as the sensors of detector systems 35 and 36 shown in FIG. 1. The sensors themselves may be externally powered or self powered and operate continuously. Ex-core Neutron Detector System 200 or Few-String Fixed In-Core Detector System 201 also operates continuously to process the raw sensor signals and to pass the conditioned output signals to the Axial Power Distribution Synthesizer 301. The methodolgy and apparatus of both detector systems 200 and 201 are well known in the art and is not included in this disclosure. The conditioned detector signals generated by ExCore Neutron Detector System 200 or Few-String Fixed In-Core Detector System 201, together with readings of control rod positions, cold-leg temperature and reactor coolant system pressure derived from Rod Control System 21, cold-leg temperature detector 38 and Pressure system 51-53, respectively, (which detectors and systems are generically grouped into Other Plant Process Monitors 300), are input into Axial Power Distribution Synthesizer 301. If the nuclear detectors in use in a particular application of the present invention are of a type referred to in the art as "two-section ex-core detectors or, equivalently, "long ion chambers," a small number of core-exit temperature values computed in Core-Exit Thermocouple System 71 may also be input to Axial Power Distribution Synthesizer 301. Axial Power Distribution Synthesizer 301 combines the input signals with the coefficient values stored in Library of Axial Expansion Coefficients 87 to generate a pointwise representation of the axial nuclear power distribution in the region of the reactor core nearest to, or including, the nuclear detectors. Methodologies appropriate for use in Axial Power Distribution 301 are well known in the art. In addition, a method for using the outputs from two-section ex-core detector systems to provide the axial power distribution is disclosed in a patent application assigned to the present assignee of this patent application. The application, having Ser. No. 850,195 and being entitled "Axial Power Distribution Monitor And Display Using Outputs From Ex-Core Detectors And Thermocouples," was filed on Apr. 10, 1986 and is incorporated by reference herein. The values representing the axial power distribution are supplied to Radial Power Distribution Expander 302 in which an estimated representation of the current three dimensional power distribution is generated by applying the simple relationship: EQU r.sup.est.sub.i,j,k =q.sub.k * p.sub.i,j /p.sub.i',j' (3) where: r.sup.est.sub.i,j,k is the "first cut" local nuclear power level at the radial location (i,j) and the axial location (k) in the reactor core; q.sub.k is the value at axial location (k) of the axial power distribution generated by Axial Power Distribution Synthesizer K; and p.sub.i,j /p.sub.i',j' is the radial power ratio relating local power level at radial location (i,j) to local power level at the reference radial location (i',j') at which the axial power distribution q.sub.k is defined. Values appropriate to all radial locations of interest and to all axial core regions of interest are stored in Library of Radial Power Ratios 88. An axial core region may be defined by a unique arrangement of inserted control rods, whence control rod position information derived from Other Plant Process Monitors 300 is supplied as input to Radial Power Distribution Expander 302, or by the nuclear characteristics of the reactor fuel which may vary axially in current reactor core designs. The r.sup.est.sub.i,j,k values which describe the nominal three dimensional power distribution in the reactor core are passed to Power Distribution F.sub..DELTA.H Calculator 303. In Power Distribution F.sub..DELTA.H Calculator 303, values are determined for the relative enthalpy rise parameter F.sup.PD.sub..DELTA.Hi,j based on the nominal or estimated three dimensional power distribution values at each fuel assembly location (i,j) by application of the relations: EQU F.sup.PD.sub..DELTA.Hi,j ={.SIGMA..sup.K.sub.k=1 r.sup.est.sub.i,j,k }/I (4) where: EQU I={.SIGMA..sup.M.sub.m=1 .SIGMA..sup.N.sub.n=1 .SIGMA..sup.K.sub.k=1 r.sup.est.sub.m,n,k }/N.sub.TFA (5) and N.sub.TFA is the total number of fuel assemblies in the core. The values of the power distribution enthalpy rise parameter F.sup.PD.sub..DELTA.H at the fuel assembly locations at which active core-exit thermocouples are available are transmitted to F.sub..DELTA.H Deviation Calculator 304. Concurrently, or in close chronological sequence, values of the core-exit coolant temperatures generated by Core-Exit Thermocouple System 71 are passed to Thermocouple F.sub..DELTA.H Calculator 91, where values of the thermocouple relative enthalpy rise F.sup.TC.sub..DELTA.H are calculated as described above. The resulting values of the thermocouple relative enthalpy rise parameter F.sup.TC.sub..DELTA.H at each active core-exit thermocouple location are also transmitted to F.sub..DELTA.H Deviation Calculator 304. In F.sub..DELTA.H Deviation Calculator 304, values of the fractional deviations of the power distribution relative enthalpy rise parameter from the corresponding thermocouple relative enthalpy rise parameter are calculated according to the relation: EQU .sigma.F.sub..DELTA.He ={F.sup.PD.sub..DELTA.He -F.sup.TC.sub..DELTA.He }/F.sup.TC.sub..DELTA.He (6) where e identifies the active core-exit thermocouple locations. These fractional deviation values .sigma.F.sub..DELTA.He are input to F.sub..DELTA.H Deviation Interpolator/Extrapolator 93 where, by the methods described above, a complete two dimensional array of values of the fractional deviation of the power distribution relative enthalpy rise parameter from either measured or hypothetical thermocouple relative enthalpy rise parameter values at all fuel assembly locations is developed. The interpolated/extrapolated array of fractional deviation values, .sigma.F.sub..DELTA.Hij, is then recombined with the nominal three dimensional power distribution values r.sup.est.sub.i,j,k in Power Distribution Adjuster 305. The recombination takes the form of: EQU r.sup.adjusted.sub.i,j,k =r.sup.est.sub.i,j,k /(1+.sigma.F.sub..DELTA.Hi,j) (7) and has the effect of combining ex-core measurements of the radial components of the current actual core power distribution with ex-core or very limited incore measurements of the axial components of the actual power distribution through the agency of deviations from a measured (via periodic flux maps) or analytically calculated reference three dimensional core power distribution. The adjusted three dimensional power distribution values r.sup.adjusted.sub.i,j,k are input to Power Distribution Normalizer 306. In Power Distribution Normalizer 306, a value of the total core output power, Q.sup.nuclear, is computed by application of the relation: EQU Q.sup.nuclear= C.sub.1 {.SIGMA..sup.I.sub.i=1 .SIGMA..sup.J.sub.j=1 .SIGMA..sup.K.sub.k=1 r.sup.adjusted.sub.i,j,k } (8) where C.sub.1 is a constant that takes into account the units in which the values of the adjusted power distribution are expressed, the fraction of the total core volume associated with typical mesh point i,j,k and the units in which total nuclear power Q.sup.nuclear, is desired. Next, a value of the total thermal power output of the core, Q.sup.thermal, is determined using the relationship: EQU Q.sup.thermal=Q.sub.rated full power * * {.DELTA.h.sub.measured /.DELTA.h.sub.full-power-reference} (9) where .DELTA.h.sub.measured is given by: EQU .DELTA.h.sub.measured =h.sub.out (T.sub.hot-leg,P)-h.sub.in (T.sub.cold-leg,P) (10) where: h.sub.out (T.sub.hot-leg,P) is the enthalpy of the coolant at average hot-leg temperature T.sub.hot-leg and reactor coolant system pressure P; h.sub.in (T.sub.cold-leg,P) is similarly defined; and .DELTA.h.sub.full-power-reference is the reactor vessel enthalpy rise corresponding to operation of the reactor at rated full power, and is established by conventional calorimetric techniques during startup and early operation of the plant. Finally, values of the synthesized three dimensional core power distribution, r.sup.synthesized.sub.i,j,k are calculated in Power Distribution Normalizer 306 in the form: EQU r.sup.synthesized.sub.i,j,k =r.sup.adjusted.sub.i,j,k * {Q.sup.thermal /Q.sup.nuclear } (11) The final pointwise values of the synthesized three dimensional core power distribution r.sup.synthesized.sub.i,j,k are made available by Power Distribution Normalizer 306 to Alarm Assessor 307 which performs functions very similar to Alarm Assessor 96 and transmits appropriate signals to Control Room Alarm 97. The final pointwise values are also transmitted to Permanent Record Device 308 which is typically a magnetic tape unit or a line printer. Further, the final pointwise values are input to Three Dimensional Burnup Accumulator 309 wherein current values of the three dimensional burnup distribution are periodically updated using the relation: EQU BU.sub.i,j,k (t)=BU.sub.i,j,k (t-.DELTA.t)++r.sup.synthesized.sub.i,j,k * .DELTA.t * C.sub.2 (12) in which .DELTA.t is the length of time interval since the last update and C.sub.2 is a constant that is used to convert the units of r and t to the units of BU. Three Dimensional Burnup Accumulator 309 also communicates periodically with Permanent Record Device 308 to permanently record current burnup distributions. The synthesized three dimensional core power distribution values are also passed to Pseudo Three Dimensional Graphics Generator 310. Pseudo Three Dimensional Graphics Generator 310 functions in a manner similar to Two Dimensional Graphics Generator 95 to develop, at a series of successive elevations in the core, color coded representations of the fractional differences or deviations between the current synthesized power distribution values and the corresponding values in a reference three dimensional pointwise power distribution, as derived either from a reference flux map or from analytical calculations. The pointwise values of the reference three dimensional core power distribution are stored in Library of Reference Three Dimensional Power Distribution Values 90. Typically, the color coding is identical to that used in the Two Dimensional Graphics Generator 95, although a different color coding scheme could be used, if warranted. FIG. 10 provides an illustration of a "stacked planes" pseudo three dimensional graphics display that appears on Graphics Monitor 100 after control rods ("Five Rod D Bank") have been inserted into the core in response to a power load decrease. In FIG. 10, axis 601 is taken along the axial direction in the core and curve 602 represents the axial power distribution in the core. Planes 700, 750, 800, and 850 represent radial power distributions in the core taken at axial heights indicated by points 605, 610, 615 and 620, respectively, along axis 601. In accordance with the inventive system, (1) in plane 700, regions 701-704 are light orange and region 705 is yellow--indicating that, on the average, the power level in the bottom quarter of the core is somewhat higher than that recorded in the reference three dimensional core power distribution stored in Library of Reference Three Dimensional Power Distribution Values 90; (2) in plane 750, regions 751-754 are light orange, region 755 is green and region 756 is yellow--indicating again that the current core power distribution is shifted slightly toward the bottom half of the core when compared to the reference three dimensional core power distribution (this shift is expected since control rods have been inserted at the top of the core); (3) in plane 800, regions 801-804 are yellow and region 805 is green--indicating only a slight increase in local power in the upper middle quarter of the core due to control rod insertion (the light orange regions in the lower two quarters correspond to yellow zones in the upper middle quarter and are probably the result of radial power redistribution in response to the initiating decrease in total core power level); and (4) in plane 850, 851-855 are light blue and surround regions of dark blue, region 856 is green, and regions 861-864 are yellow--wherein the insertion of the "Five Rod D Bank" control rods cause the similar regions of locally depressed power in the top quarter of the core along with associated regions of slightly higher power (the yellow regions 861-864) in areas radially remote from the control rod X-Y locations (The similarities among the five locally depressed regions of core power in the top quarter of the core show clearly that all control rods in the "Five Rod D Bank" have remained mutually aligned, i.e., no control rod drive malfunction has occurred). Again, other displays based on the same "stacked planes" principle and showing, for example, color coded representations of absolute local power levels can readily be constructed from the synthesized three dimensional power distribution values generated by this second aspect of the present invention. While specific embodiments of the invention have been described in detail, it will be appreciated by those skilled in the art that various modifications and alternatives to those details could be developed in light of the overall teachings of the disclosure. Accordingly, the particular arrangements disclosed are meant to be illustrative only and not limiting as to the scope of the invention which is to be given the full breadth of the appended claims and any and all equivalents thereof.
047568674	description	DESCRIPTION OF PREFERRED EMBODIMENT FIGS. 1 and 1a show the lower part of the pole 1 of great length, making it possible to install a measuring apparatus, designated as a whole by the reference 2, on the lower core plate 3 of a pressurized-water nuclear reactor in the region of a guide bush 4 allowing a glove finger to pass throguh towards a fuel assembly. In the drawings, the device has been shown in the operating position on the lower core plate 3, the reactor being shut down, the core assemblies being removed and the pool and reactor vessel being filled with water. The handling pole 1, whose upper part (not shown) is located above the reactor pool and which is of a length greater than ten meters, makes it possible to lower the measuring apparatus 2 to the level of the lower core plate and move it along, for example using the reactor loading machine to carry out the handling operation. The measuring apparatus 2 comprises a supporting structure 5 connected to the lower part of the pole 1 by means of a suspension device 6 and a measuring assembly 7 carried by the supporting structure 5. The suspension device 6 comprises an upper plate 8 for connection to the pole 1, a middle plate 10 for connection to the supporting structure 5, three suspension arms 9 connecting the plates 8 and 10, and three columns 12 connecting the intermediate plate 10 to the plane annular base 5 forming the supporting structure of the measuring apparatus. Furthermore, the intermediate plate 10 carries, in its central part, a television camera 13 making it possible to display the measuring zone. The annular base 5 is fastened by means of screws 14 to the lower end of the suspension columns 12 which are themselves fastened, in their upper part, to the intermediate plate 10 by means of screws 15. The annular base 5 is integral with two positioning studs, such as 16, projecting downwards and of a shape corresponding to the shape of the water passage orifices 17 extending through the lower core plate 3. The annular base 5 also carries a thrust assembly 18 consisting of a support 19, a thrust arm 20 and an actuating finger 21. The thrust arm 20 is mounted in an articulated manner on the support 19 about a horizontal pivot pin 22 and, at one of its ends which is bent, carries a bearing roller 23 mounted so as to be rotatable about a horizontal pivot pin. The other end of the arm 20 bears by menas of a spring 24 on a board 25 fastened to the support 19. The finger 21 comprises an outer bush 21a and a pusher 21b mounted slideably in the bush 21a. A spring 27 is interposed between the bush 21a and the pusher 21b. The bush 21a is engaged and fitted slideably in an orifice passing through the support 19 and has an upper retaining edge 28 which allows the bush 21a to come to rest on the support 19 when the measuring apparatus 2 is lifted above the plate 3 by means of the pole 1. In this position of rest, the finger 21 then projects below the lower surface of the base 5 intended to come into contact with the upper surface of the lower core plate 3. FIG. 1 shows the device in the operating position on the lower core plate 3, and in this operating position the finger 21 is up against the thrust arm 20 and is brought up against the latter during the installation of the measuring apparatus 2 on the lower core plate 3. The finger 21 allows the arm 20 to tilt about the pivot pin 22, so as to move its bearing roller 23 to the right (in FIG. 1). As can be seen in FIGS. 1a and 2, the measuring assembly 7 has a plane stage 30 mounted on the base 5, so as to be paralell to its plane faces, by means of springs 31 fitted round the lower part of the columns 12 and interposed between the upper surface of the base 5 and a washer 32 bearing under the lower face of the stage 30. The upper face of the stage 30 abuts against a stop element 33 of the column 12. In line with each of the columns 12, the stage 30 has a circular orifice 35 of a diameter substantially greater than the diameter of the column 12. The stage 30 can thus move parallel to the plane of the faces of the base 5 in all directions, either in a translational movement or in a rotational movement, by an amount determined by the play between the orifice 35 in the stage and the corresponding column 12. Furthermore, the stage can be inclined about horizontal axes, in order to make it easier to install the measuring apparatus. The stage 30 has a V-shaped orifice which can be seen in FIG. 2 and which communicates with a circular central orifice, into which is introduced a tubular sleeve 36 arranged with its axis perpendicular to the stage 30. The sleeve 36 is integral with a support 37, in which is mounted a measuring tracer with a movable rod 38, with which a movement measuring means, to be described later, is associated. The tubular sleeve 36 has an inside diameter slightly greater than the maximum outside diameter allowed to the various guide bushes 4, on which the measurement is made. The support 37 and consequently the tubular sleeve 36 are fastened rigidly to the stage 30 by means of screws 39. In its position shown in FIG. 1, the thrust arm 20 bears by means of its roller 23 on the outer surface of the tubular sleeve 36, in a zone substantially diametrically opposite the position of the tracer 38. The movable rod 40 of the tracer 38 is pushed towards the inside of the sleeve 36 by a spring 41, in such a way that the end of this rod 40 projects slightly inside the sleeve in its position of rest. In the operating position demonstrated in FIG. 1, the tubular sleeve 36 is slipped onto the outer surface of the guide bush 4, on which the diameter measurement is made, and the end of the rod 40 is in contact with the outer surface of the bush 4. The inner surface of the tubular sleeve 36 is machined to form two slightly projecting fixed stops arranged at 120 relative to the rod of the tracer 40 round the axis of the sleeve 36. In this way, when the thrust arm 20 bears by means of its roller 23 on the bush 36, the entire measuring assembly movable relative to the base 5 moves up to the moment when the fixed stops come into contact with the bush 4. At this moment, the position of the rod 40 of the tracer 38, which is up against the guide bush 4 at a point opposite the fixed stops, depends only on the outside diameter of the bush 4 or on the difference between this diameter and one or more reference values. The position of the rod 40 must therefore be determined accurately, and, to achieve this, the end of this rod opposite the end in contact with the bush 4 comes up against a ball 45 which is itself in contact with a sloping surface 46 integral with the support 37. This assembly makes it possible to transfer, at an angle of 90.degree. and with a ratio of 1, the movements of the rod 40 which are thus transmitted to a vertical rod 50 mounted in the support 37 and returned by a spring 49. The rod 50, in its upper part, is integral with an actuating plate 48, of which the movements of position representing the movements or position of the rod 40 of the tracer can be measured in two different ways. A rocker-type comparator with a dial 52 comprises a rod 53 located just above the plate 48 and a dial located in the viewing field of the camera 13. The movements or the position of the rod 40 can thus be read directly. An electronic movement sensor 54 is also arranged above the plate 48, with its rod 55 vertical and in the extension of the rod 50. The electronic sensor 54 is connected, by means of a cable 56, to an electronic processing box associated with a display screen making it possible to read directly the numerical values representing the position of the sensor. The mode of operation of the device will now be described in relation to the measurement of the diameter of a guide bush fastened to the lower core plate of a pressurized-water nuclear reactor, this plate being under a head of water exceeding 10 meters in the vessel of the shutdown reactor. Before the actual measuring operations, calibration is carried out using a calibrating plate having four orifices identical to the water passage orifices in the lower core plate, and two guide bushes which are perpendicular to the plate and the respective diameters of which are the minimum diameter and the maximum diameter of the guide bushes of the lower core plate, on which the measurement is made. These diameters are 31.50 and 32.00 mm, respectively. The calibrating plate also has a hole tapped in its central part, making it possible to fasten it by screwing to the end of the pole 1, in order to lower it to the bottom of the reactor pool. At the bottom of the pool, the measuring device is placed in position on the bush of minimum diameter and on the bush of maximum diameter of the calibrating plate. The corresponding values are recorded by means of the processing and display box, thus allowing direct diameter calibration under the precise operating conditions of the measuring device. To put it into operation, the measuring apparatus is lowered to the bottom of the pool by means of the handling pole 1, the upper end of which is fastened to the reactor loading machine. The apparatus is first brought into a position located 100 mm above the lower core plate 3. The operator then searches for the best possible centering by means of the video camera 13, the upper part of the guide bush 4 appearing in the center of the bore in the sleeve 36. The device then continues to be lowered, its projecting guide and centering studs 16 penetrating into the corresponding water passage orifices 17 in the lower core plate 3. The supporting structure of the device and its base 5 in particular are then centered approximately relative to the bush 4. The sleeve 36 of the movable measuring assembly of the device is then matched up with the upper part of the bush 4. To make engagement easier, the lower part of the inner surface of the sleeve 36 is widened downwards and interacts with the upper frusto-conical part of the bush 4. During the engagement and fitting of the sleeve 36 round the bush 4, the movable measuring assembly can be inclined in all directions at a small angle of deflection and can move in all the directions of the plane parallel to the plane bearing face of the base 5 which comes into contact and merges with the upper plane face of the lower core plate 3. Before it comes into contact in this way, when, during lowering, the contact face of the base 5 arrives at a distance of 5 mm from the lower core plate, the finger 21 comes into contact with the lower core plate 3 by means of its bush 21a. When the measuring device continues to be lowered, the finger 21 moves in the vertical direction, comes into contact with the arm 20 by means of its pusher 21b and causes the arm 20 to tilt in such a way that its roller 23 moves to the right (in FIG. 1). The bearing roller 23 pushes back the sleeve 36 and the movable measuring assembly, up to the moment when the fixed stops inside the sleeve 36 come into contact with the outer surface of the guide bush 4. At this moment, the position of the rod 40 of the tracer 38 represents the diameter of bush. This position is recorded by means of the mechanical sensor 52 and the electronic sensor 54, and the diameter value in millimeters can be displayed directly by means of the processing box and its display screen. For this purpose, the electronic sensor 54 is supplied with measuring current via the processing box which makes it possible to process the signals feeding the sensor 54 by means of an oscillator/demodulator module. The signals transmitted in return by the sensor 54, which is of the LVDT type (linear-displacement differential transformer), are received by the processing box which converts the electrical information of the signals into, for example, digital information representing the diameter of the guide bush or its variation in relation to a specific diameter value. The processing box also has an output making it possible to provide permanently a signal in analog form which represents the diameter or its variations and which can be utilized, for example, on a chart recorder. The information in digital form can be used not only for displaying the diameter value but also for processing it in a computer. A value representing the diameter of the guide bush can be read directly by the operator on the dial 52 by means of the video camera 13. The arrangement at 120.degree. of the fixed stops and of the rod 40 of the tracer 38 makes it possible to carry out installation and measurement under very reliable conditions. Thus, by means of the measuring device according to the invention, it becomes possible to achieve an accuracy greater than 0.01 mm and a fidelity better than 0.001 mm, under the conditions of use which have been described. During the measurements, to prevent any disturbing influence of even a slight stress transmitted to the measuring device by means of the handling pole 1, the pole continues to be lowered after the measuring device rests on the lower core plate 3. This makes it possible to release the device completely, the connection between the pole 1 and the suspension device 6 of the measuring apparatus being of the shackle type. It will be seen from the foregoing description that the measuring device is perfectly safe, completely reliable and easy to use. The elements are put in position on the lower core plate and round the guide bush automatically, from the moment when pre-positioning has been carried out by means of the video camera associated with the measuring device. The diameter of the guide bushes or the variation in this diameter can be read or recorded directly by processing, display and recording means which are completely conventional. The sleeve and the bush can be brought into contact with one another, before the measurement, by a means different from a set of two fixed stops located opposite the tracer rod. The base of the supporting structure coming to rest on the lower core plate can be connected to the handling pole by means of a device different from the suspension device with a tie and columns which has been described. Likewise, the movable measuring assembly can be mounted on this base in a different way from that described, which used the suspension columns and bearing springs allowing a flexible mounting. The movement sensors can be of a type different from those described, and it is possible to use a single type of sensor just as well as several sensors in parallel. It would also be possible to use a sensor which makes it possible to measure directly the movements of the rod of the measuring tracer, without transfer at an angle. Finally, the device according to the invention can be used for operations other than the checking of the guide bushes of the lower core plate of a pressurized-water nuclear reactor. It can be used whenever it is necessary to measure very accurately and remotely the diameter of a cylindrical element projecting relative to a fixed plate perforated with orifices in the vicinity of the cylindrical element.
	abstract	A device and a method to determine a position of a component that is moveable in a linear manner along an assigned axis are provided. A reference element that extends in a direction of the assigned axis is assigned to the component. The component may be brought into mechanical contact with the reference element. A respective piezo transducer is arranged on the reference element to generate and receive vibrations in a material used for the reference element.
	abstract	A method of manufacturing X-ray lenses which transmit X-rays which has a first step of providing a layer of liquid on the flat surface of a first substrate, a second step of arranging numerous pipe-shaped lens components in a row following an axis which extends parallel to the flat surface in the layer of liquid, and a third step of holding the pipe-shaped lens components between the surface of a second substrate having a flat surface and the flat surface of the first substrate, and filling the liquid in spaces formed by the exterior surface of the pipe-shaped lens components and the flat surface of the first substrate or the flat surface of the second substrate. The pipe-shaped lens components can be carbon nanotubes, and the liquid can be a mixture of a solvent and a lubricant such as silicon grease which has had its viscosity reduced.
060552883	claims	1. A nuclear reactor vessel having a baffle-barrel assembly for supporting fuel assemblies in a core region and for guiding fluid flowing through the core region when the reactor vessel is in service, comprising: a baffle plate defining a countersunk hole having a diameter and a smaller diameter baffle plate bolt hole extending from the countersunk hole, the baffle plate defining a slot extending through the baffle plate and outwardly from both the countersunk hole and the baffle plate bolt hole; a former plate having opposed surfaces, the former plate defining a bolt hole aligned with the baffle plate bolt hole, the former plate further defining a slot extending from the baffle plate slot and extending outwardly from the former plate bolt hole to at least one of the opposed surfaces; and a bolt for fastening the two plates together, the bolt having a head portion with an undersurface disposed in the countersunk hole and a shank extending from the undersurface of the head portion into the aligned bolt holes, the bolt head defining at least in part a fluid flow passageway external of the shank interconnecting the countersunk hole with the baffle plate bolt hole; and wherein the outwardly extending baffle plate slot extends to at least one of the opposed surfaces of the former plate. a baffle plate defining a countersink hole having a diameter and a smaller diameter baffle plate bolt hole extending from the countersunk hole; a former plate defining a bolt hole aligned with the baffle plate bolt hole; and a bolt for fastening the two plates together, the bolt having a head portion with an undersurface disposed in the countersunk hole and a shank extending from the undersurface of the bead portion into the aligned bolt holes, the bolt head entirely defining a fluid flow passageway external of the shank interconnecting the countersunk hole with the baffle plate bolt hole. a baffle plate defining a countersunk hole having a diameter and a smaller diameter baffle plate bolt hole extending from the countersunk hole; a former plate defining a bolt hole aligned with the baffle plate bolt hole; a bolt fastening the two plates together, the bolt having a head portion with an undersurface disposed in the countersunk hole and a shank extending from the undersurface of the head portion into the aligned bolt holes, the bolt head portion having a slot interconnecting the countersunk hole with the baffle plate bolt hole externally of the shank. removing the nuclear reactor vessel from service; and then machining a slot extending through the baffle plate and outwardly from both the baffle plate countersink bole and the baffle plate aligned hole and further extending the slot into the aligned former plate outwardly from the aligned former plate hole such that both the slot in the baffle plate and the communicating slot in the former plate extend to at least one of the opposed surfaces of the former plate; whereby fluid flowing through the reactor vessel when the reactor vessel is in service will wash the undersurface of the bolt head portion. crimping a bolt locking cup into the baffle plate slot. machining a second slot into the baffle plate extending through the baffle plate and outwardly from both the baffle plate countersunk hole and the baffle plate aligned hole, wherein the two slots have different dimensional lengths extending outwardly. removing the nuclear reactor vessel from service; and then replacing the existing baffle/former bolt with a baffle/former bolt having a head portion defining a slot external of the shaft interconnecting the countersunk hole and the baffle plate bolt hole; whereby fluid flowing through the reactor vessel when the reactor vessel is in service will wash the undersurface of the bolt head portion. crimping a bolt locking cup into the bolt head slot. removing the nuclear reactor vessel from service; and then replacing the existing baffle/former bolt with a baffle/former bolt having a head portion entirely defining a passageway external of the shaft interconnecting the countersunk hole and the baffle plate bolt hole; whereby fluid flowing through the reactor vessel when the reactor vessel is in service will wash the undersurface of the bolt head portion. 2. The vessel of claim 1, further comprising: a bolt locking clip protruding into the baffle plate slot. 3. The vessel of claim 1, wherein the bolt defines a second fluid flow passageway extending through the head portion and internally through the shank in fluid flow communication with the countersunk hole and the aligned bolt holes. 4. A nuclear reactor vessel having a baffle-barrel assembly for supporting fuel assemblies in a core region and for guiding fluid flowing through the core region when the reactor vessel is in service, comprising: 5. A nuclear reactor vessel having a baffle-barrel assembly for supporting fuel assemblies in a core region and for guiding fluid flowing through the core region when the reactor vessel is in service, comprising: 6. The vessel of claim 5, further comprising: a bolt locking cup protruding into the slot. 7. The vessel of claim 5, wherein the bolt defines a fluid flow passageway extending through the head portion and internally through the shank in fluid flow communication with the countersunk hole and the aligned bolt holes. 8. A method of backfitting a nuclear reactor vessel having a baffle plate fastened to a former plate by an existing baffle/former bolt, the bolt having a head portion with an undersurface disposed in a countersunk hole in the baffle plate with a shank extending from the undersurface through aligned holes in the baffle plate and a former plate, the former plate having opposed surfaces, comprising the steps of: 9. The method of claim 8, including the step of: 10. The method of claim 13, including the additional step of: 11. The method of claim 8, wherein the former plate slot extends from the aligned bolt hole to both opposed surfaces. 12. A method of backfitting a nuclear reactor vessel having a baffle plate fastened to a former plate by an existing baffle/former bolt, the bolt having a head portion with an undersurface disposed in a countersunk hole in the baffle plate with a shank extending from the undersurface through aligned holes in the baffle plate and a former plate, comprising the steps of: 13. The method of claim 12, including the step of: 14. A method of backfitting a nuclear reactor vessel having a baffle plate fastened to a former plate by an existing baffle/former bolt, the bolt having a head portion with an undersurface disposed in a countersunk hole in the baffle plate with a shank extending from the undersurface through aligned holes in the baffle plate and a former plate, comprising the steps of:
	abstract	A planning apparatus (70) determines irradiation parameter data (67) for a charged particle irradiation system (1), which radiates charged particles generated by an ion source (2) to a target (80) by accelerating the charged particles by means of a linear accelerator (4) and a synchrotron (5). The planning apparatus is provided with: a planning program (73), which determines the irradiation parameter data (67) with respect to one target (80) by combining charged particles of a plurality of kinds of ion species; and a CPU (71) for executing the planning program. Consequently, the irradiation planning apparatus capable of performing irradiation with desirable dose distribution with respect to the target, the irradiation planning program, an irradiation plan determining method, and the charged particle irradiation system are provided.
	summary
047284897	abstract	A fuel element support grid for supporting a plurality of nuclear fuel elements intermediate their ends has at least some of the pairs of intersecting and slottedly interlocked strips including pairs of intersecting integral fluid flow directing vanes along at least one adjacent edge of each of the strips of the pair. Welds attach the pair of vanes to each other thereby providing welded attachment of the strips. The welds may be at the intersection of the vanes remote from their areas of integral attachment to their respective strips or they may be adjacent to their areas of integral attachment to their respective strips.
	claims	1. A system of non-uniformity pattern identification, the system comprising:a storage device capable of storing a plurality of theoretical patterns and a plurality of measurements, in which each measurement corresponds to a region on a wafer; anda processing unit configured to acquire the theoretical patterns and the measurements on at least two wafers, calculate pattern scores for the respective theoretical patterns of each wafer according to the measurements, in which each pattern score represents the extent of similarity between one of the theoretical patterns and the measurements on one of the wafers, and groups at least two of the theoretical patterns into at least one factor according to the pattern scores to identify a non-uniformity pattern for the wafers. 2. The system of claim 1 wherein the theoretical patterns comprise a uniformity pattern and a plurality of non-uniformity patterns. 3. The system of claim 1 wherein the theoretical patterns are implemented in a matrix, a two-dimensional array, a linked list or a tree. 4. The system of claim 1 wherein the region covers one or more dies on the wafer, or covers a portion of one die. 5. The system of claim 1 wherein the measurements are electrical measurements or physical measurements. 6. The system of claim 1 wherein the pattern scores are calculated by a correlation analysis algorithm or a data classification method according to the measurements. 7. The system of claim 1 wherein the pattern scores are calculated by an equation: MT m × m × [ W1 ⋮ ⋮ ⋮ Wm ] × 1 L = [ P1 P2 ⋮ ⋮ Pm ] ,wherein MTm×m represents the m-by-m matrix for m theoretical patterns, W1 to Wm represent measurements individually occurring in the respective regions, L represents an individual standardization factor, which is the square root of the sum of the square of the cell values for each row 1 to m, and P1 to Pm represent the pattern scores. 8. The system of claim 1 wherein the theoretical patterns are grouped into the factor using a principal component analysis (PCA) or a data clustering algorithm. 9. The system of claim 1 wherein the factor has highest explanability. 10. The system of claim 1 wherein the measurements are acquired during a wafer acceptance test (WAT) or an in-line processing measurement. 11. The system of claim 1 wherein the processing unit outputs a graph corresponding to the factor to an output device. 12. The system of claim 11 wherein the graph comprises a contour, a box plot chart or a histogram. 13. A method of non-uniformity pattern identification, the method comprising using a computer to perform the steps of:acquiring a plurality of theoretical patterns;acquiring a plurality of measurements on at least two wafers, in which each measurement corresponds to a region on one wafer;calculating pattern scores for the respective theoretical patterns of each wafer according to the measurements, in which each pattern score represents the extent of similarity between one of the theoretical patterns and the measurements in one of the wafers; andgrouping at least two of the theoretical patterns into at least one factor according to the pattern scores to identify a non-uniformity pattern for the wafers. 14. The method of claim 13 wherein the theoretical patterns comprise a uniformity pattern and a plurality of non-uniformity patterns. 15. The method of claim 13 wherein the theoretical patterns are implemented in a matrix, a two-dimensional array, a linked list or a tree. 16. The method of claim 13 wherein the region covers one or more dies on the wafer, or covers a portion of one die. 17. The method of claim 13 wherein the measurements are electrical measurements or physical measurements. 18. The method of claim 13 wherein the pattern scores are calculated by a correlation analysis algorithm or a data classification method according to the measurements. 19. The method of claim 13 wherein the pattern scores are calculated by an equation: MT m × m × [ W1 ⋮ ⋮ ⋮ Wm ] × 1 L = [ P1 P2 ⋮ ⋮ Pm ] ,wherein MTm×m represents the m-by-m matrix for m theoretical patterns, W1 to Wm represent measurements individually occurring in the respective regions, L represents an individual standardization factor, which is the square root of the sum of the square of the cell values for each row 1 to m, and P1 to Pm represent the pattern scores. 20. The method of claim 13 wherein the theoretical patterns are grouped into factors using a principal component analysis (PCA) or a data clustering algorithm. 21. The method of claim 13 wherein the factor has highest explanability. 22. The method of claim 13 wherein the measurements are acquired during a wafer acceptance test (WAT) or an in-line processing measurement. 23. The method of claim 13 further comprising a step of outputting a graph corresponding to the factor to an output device. 24. The method of claim 23 wherein the graph comprises a contour, a box plot chart or a histogram. 25. A machine-readable storage medium for storing a computer program which when executed performs a method of non-uniformity pattern identification, the method comprising the steps of:acquiring a plurality of theoretical patterns;acquiring a plurality of measurements on at least two wafers, in which each measurement corresponds to a region on one wafer;calculating pattern scores for the respective theoretical patterns of each wafer according to the measurements, in which each pattern score represents the extent of similarity between one of the theoretical patterns and the measurements in one of the wafers; andgrouping at least two of the theoretical patterns into at least one factor according to the pattern scores to identify a non-uniformity pattern for the wafers.
	abstract	With a detector system for the specimen chamber of a scanning electron microscope, signals are simultaneously detected in transmission which signals correspond to a light field contrast and a dark field contrast. The detector system (14) includes four detectors (15 to 18) in a plane (25) between which an aperture (19) for free access of electrons is located. Behind the aperture (19), a further detector (27) is arranged in a second plane (26). The detectors are preferably diodes. The detectors (15, 16, 17, 18) in the first plane (25), which is closer to the specimen, serve to generate signals which correspond to a dark field contrast. The further detector (27), more distant from the specimen, detects signals corresponding to a light field contrast. Large dead spaces, which are not sensitive to electrons, between the diodes and around the aperture (19), can be avoided by the offset arrangement of four diodes (15, 16, 17, 18) in the first plane (25).
H00005088	summary	BACKGROUND OF THE INVENTION The invention described herein relates generally to inertial confinement fusion (ICF), and more particularly to methods and apparatus for reducing the input energy requirement for driving ICF targets. The avowed purpose of ICF is to produce relatively tiny but powerful thermonuclear explosions by imploding small DT-filled targets to ignition conditions. The very energetic thermonuclear products released from the explosions are intended to be used to produce electricity, to provide high-energy x-rays and neutrons for important scientific experiments, and to accomplish many other useful and beneficial goals. Even though ICF targets are driven by presently existing means to provide modest quantities of thermonuclear energy, the full realization of the potential of ICF will only be reached with the routine attainment of thermonuclear yields in excess of about 0.1 ton, TNT equivalent. In operation, ICF targets are presently set and kept in motion by either one or the other of two distinctly different types of driver. As described by Nuckolls et al in Nature 239, 139 (1972), ICF targets may be driven by lasers. Alternatively, as set forth by Clauser in Phys. Rev. Lett. 35, 848 (1975), ICF targets may be driven by ion beams, where the ions may include electrons and charged atoms or groups of atoms. The gain of an ICF target is defined as the ratio of the amount of thermonuclear energy released by the target, to the amount of energy provided by the driver. The gains of all presently existing ICF target systems, be they laser or ion beam driven, are considerably less than unity. Clearly, for any ICF target system to be practically viable, its gain will have to be well in excess of unity. The gain of any ICF target system is functional of many parameters, such as uniformity of target illumination, driving pulse shape, photon energy in the case of laser drivers, and particle species and energy in the case of ion beam drivers. Additionally, with all other parameters held fixed, the gain of any ICF target system is usually increased by increasing the amount of energy provided by the driver, at least within the range of driver energies that are presently available. It is finally pointed out that the best and most efficient ICF target drivers that presently exist, even though they are incapable of providing gains of or in excess of unity, are huge and extremely complicated, extradordinarily expensive pieces of scientific apparatus--with sizes measured in hundreds of feet and costs measured in hundreds of millions of dollars. It is, therefore, apparent that any methods or apparatus for increasing ICF target gain, while keeping the amount of available driver energy fixed, would be of extraordinary importance to making the goals of ICF more attainable. SUMMARY OF THE INVENTION It is, therefore, an object of the invention to provide method and apparatus, operative at fixed driver energy, for increasing the gain of an ICF target system. Additional objects, advantages and novel features of the invention will be set forth in part in the description which follows, and in part will become apparent to those skilled in the art upon examination of the following or may be learned by practice of the invention. The objects and advantages of the invention may be realized and attained by means of the instrumentalities and combinations particularly pointed out in the appended claims. To achieve the foregoing and other objects and in accordance with the purpose of the present invention, as embodied and broadly described herein, the method and apparatus of this invention comprises driving the implosion of an ICF target in two phases, with the target being driven in each of the phases with a separate driver. The ICF target comprises at least a hollow spherical ablator surroundingly disposed around a quantity of fusion fuel. Of course, the target may additionally comprise other elements, as appropriate. In driving the target, the ablator is first compressed to higher density by laser beams. After the ablator has been thus compressed, ion beams are then used to deliver energy into the compressed ablator. This direct energy deposition causes the ablator to implode, and compress the quantity of fusion fuel to conditions wherein fusion reactions occur. In another embodiment of the invention, that is quite similar to the embodiment just described, instead of using laser beams to initially compress the ablator to higher density, this function is performed by a first quantity of ion beams. The subsequent delivery of energy into the compressed ablator is accomplished by a second quantity of ion beams, with the implosion and compression of the quantity of fusion fuel remaining as stated above. The methodology of the invention, then, comprises compressing the ablator of an ICF target to higher density with laser beams, or, in another embodiment of the invention, with a first quantity of ion beams. The next step comprises delivering energy into the compressed ablator with an entirely independent quantity of ion beams. This implodes and compresses the quantity of fusion fuel within the ICF target to conditions wherein fusion reactions occur. Since the apparatus and methods just described allow the gain of an ICF target to be increased, while the total amount of available driver energy is kept fixed, the present invention provides the benefits and advantages attendant upon making the ultimate goals of ICF more attainable.
	abstract	A device for creating an environment in which fusion can occur is provided. In its most basic embodiment, the present invention comprises two opposing cathodes separated from each other by a gap. An anode is positioned outside of the gap on a horizontal plane from the vertically positioned cathodes. This cathode and anode structure is positioned within a chamber with a vacuum drawn. Into the chamber, a quantity of fuel such as hydrogen, deuterium, and/or tritium fuel may be introduced. Upon application of a current to the system, ions will be retained in orbit about the cathodes, creating a plasma.
	claims	1. An ion implanting apparatus, comprising:a member comprising a through hole through which an ion beam is passed to form a beam geometry,wherein at least an inner surface of said through hole of said member comprises an unoriented poly-crystalline structure. 2. The ion implanting apparatus according to claim 1, wherein at least said inner surface of said through hole of said member is coated with a thermal spraying film. 3. The ion implanting apparatus according to claim 1, wherein said thermal spraying film comprises silicon. 4. The ion implanting apparatus according to claim 1, wherein said thermal spraying film comprises tungsten. 5. The ion implanting apparatus according to claim 2, wherein said thermal spraying film has a thickness of about 150 μm. 6. The ion implanting apparatus according to claim 2, wherein said thermal spraying film comprises said unoriented poly-crystalline structure. 7. The ion implanting apparatus according to claim 2, wherein said thermal spraying film comprises a porous film. 8. The ion implanting apparatus according to claim 1, wherein at least the inner surface of said through hole of said member is coated with a coating film having an unoriented poly-crystalline structure. 9. The ion implanting apparatus according to claim 1, wherein said member comprises an oriented poly-crystalline structure, andwherein at least the inner surface of said through hole of said member is coated with a coating film having an unoriented poly-crystalline structure. 10. The ion implanting apparatus according to claim 1, wherein said member comprises carbon. 11. An ion implanting apparatus, comprising:a member comprising a through hole through which an ion beam is passed to form a beam geometry,wherein at least an inner surface of said through hole of said member is porous. 12. The ion implanting apparatus according to claim 11, wherein at least the inner surface of said through hole of said member is coated with a porous film. 13. The ion implanting apparatus according to claim 12, wherein said porous film comprises:a surface comprising a plurality of concave portions; andan interior comprising a plurality of pores,wherein at least a portion of said plurality of concave portions connect with at least a portion of said plurality of pores, andwherein at least a portion of said plurality of pores mutually connect. 14. The ion implanting apparatus according to claim 13, wherein said plurality of concave portions and said plurality of pores are not influential in forming said beam geometry, and are formed to have dimensions that are adopted for adsorbing ion species. 15. The ion implanting apparatus according to claim 13, wherein said plurality of concave portions have a thickness less than or equal to 5 μm. 16. The ion implanting apparatus according to claim 12, wherein said porous film comprises a coating film. 17. The ion implanting apparatus according to claim 16, wherein said coating film comprises a thermal spraying film. 18. The ion implanting apparatus according to claim 17, wherein said thermal spraying film comprises silicon. 19. The ion implanting apparatus according to claim 17, wherein said thermal spraying film comprises tungsten. 20. The ion implanting apparatus according to claim 11, wherein at least an inner surface of said through hole of said member comprises an unoriented poly-crystalline structure.
	description	The present application claims priority benefits under 35 U.S.C. §119 to Korean Patent Application No. 10-2010-0006471, filed Jan. 25, 2010. 1. Field of the Invention The present invention relates to a hold-down spring unit for top nozzles of nuclear fuel assemblies which are used in nuclear reactors. The hold-down spring unit has an improved hold-down performance to prevent the nuclear fuel assembly from lifting up. The present invention, also, relates to a top nozzle for nuclear fuel assemblies which employ the hold-down spring unit. 2. Description of the Related Art As is well known to those skilled in the art, a nuclear reactor is a device in which a fission chain reaction of fissionable materials is controlled for the purpose of generating heat, producing radioactive isotopes and plutonium, or forming a radiation field. Generally, in light-water reactor nuclear power plants, enriched uranium which is increased in the ratio of uranium-235 to 2% through 5%, is used. To process enriched uranium into nuclear fuel to be used in nuclear reactors, a forming process is conducted by which uranium is formed into a cylindrical pellet having a weight of about 5 g. Several hundreds of these pellets are retained into a bundle and inserted into a zirconium tube under vacuum conditions. A spring and a helium gas are supplied into the tube and a cover is welded and sealed onto the tube, thus completing a fuel rod. A plurality of fuel rods constitutes a nuclear fuel assembly and is burned in a nuclear reactor by nuclear reaction. FIG. 1 is a front view showing a conventional nuclear fuel assembly. As shown in FIG. 1, the nuclear fuel assembly includes a plurality of support grids 10 through which fuel rods (not shown) are inserted, and a plurality of guide thimbles 15, which are coupled to the support grids 10. The nuclear fuel assembly further includes a top nozzle 30, which is coupled to the upper ends of the guide thimbles 15, a bottom nozzle 16, which is coupled to the lower ends of the guide thimbles 15, and the fuel rods (not shown), which are supported by springs and dimples that are formed in the support grids 10. To assemble the nuclear fuel assembly having the above-mentioned construction, lacquer is applied to the surfaces of the fuel rods to prevent the fuel rods from being scratched and to prevent springs provided in the support grids 10 from being damaged. Thereafter, the fuel rods are inserted through the support grids 10 and then the top nozzles 30 and bottom nozzles 16 are coupled to the fuel rods, thus completing the assembly of the nuclear fuel assembly. The assembled nuclear fuel assembly is tested for distances between the fuel rods, distortion, dimensions including the length, etc. after the lacquer is removed. When the results of the test are normal, the nuclear fuel assembly is installed in a core of a nuclear reactor in which nuclear fission is produced, as disclosed in U.S. Pat. No. 5,213,757. In the nuclear fuel assembly installed in the core, a hydraulic uplift force which is generated by the flow of coolant during the operation of the nuclear reactor is applied to the top nozzles 30 and bottom nozzles 16. Hereby, the nuclear fuel assembly is lifted up or vibrated. Furthermore, thermal expansion attributable to an increase in temperature, irradiation growth of the nuclear fuel tube as a result of neutron irradiation for a long period of time, or axial length variation caused by creep may be induced. Therefore, the top nozzle 30 is configured to ensure the mechanical and structural stability of the nuclear fuel assembly with respect to axial movement or axial length variation of the nuclear fuel assembly. FIG. 2 is a perspective view of the top nozzle 30 according to a conventional technique. As shown in FIG. 2, the top nozzle 30 includes a plurality of spring clamps 31 which support hold-down spring units 32. Spring insert holes 31a are formed in each spring clamp 31. The ends of the hold-down spring units 32 are inserted into corresponding spring insert holes 31a. A fastening pin hole 32a″ is vertically formed through the end of each hold-down spring unit 32 which is inserted into the corresponding spring insert hole 31a. Each hold-down spring unit 32 includes a first spring 32a having a first neck part 32a′, a second spring 32b and a third spring 32c which are coupled to the first neck part 32a′. The hold-down spring unit 32 is configured such that the first, second and third springs 32a, 32b and 32c are stacked on top of one another. To couple the hold-down spring unit 32 to the top nozzle 30, a spring junction end of the hold-down spring unit 32 which is opposite the first neck part 32a′ is inserted into the corresponding spring insert hole 31a in the horizontal direction. Thereafter, a fastening pin 33 is inserted into the corresponding fastening pin hole 33′ of the spring clamp 31 and a fastening pin hole 32a″ of the hold-down spring unit 32 in the vertical direction. Thereby, the hold-down spring unit 32 is fastened to the top nozzle 30. Here, to prevent the fastening pin 33 from being removed, the fastening pin 33 is welded to an upper surface of the spring clamp 31. As shown in FIG. 1, the top nozzle 30 having the above-mentioned construction is assembled with the elements of the nuclear fuel assembly. Subsequently, as is well known, the nuclear fuel assembly is installed in a core and disposed between an upper core plate (not shown) and a lower core plate such that the hold-down spring units 32 are supported by the lower surface of the upper core plate. As shown in FIG. 2, the hold-down spring units 32 which are provided on the top nozzle 30 provide elastic force to the nuclear fuel assembly in response to axial movement or variation in the length of the nuclear fuel assembly so as to ensure the mechanical-structural stability of the nuclear fuel assembly. The first neck part 32a′ of the first spring 32a is inserted into an insert slot 41 formed in a corresponding upper plate 40 of the top nozzle 30 in order to guide the operation of the hold-down spring unit 32 and prevent a loss of an element when the first, second or third spring 32a, 32b or 32c is damaged. FIG. 3 is a graph showing the characteristic curve of the hold-down spring unit 32 according to the conventional technique. As shown in FIG. 3, the hold-down spring unit 32 according to the conventional technique has the hold-down margin such that the spring force is greater than the demand hold-down force in the entire operating section and the gradient of the graph showing the spring force as a function of displacement is constant. In other words, in the hold-down spring unit 32 mounted to the top nozzle 30 according to the conventional technique, because the first, second and third springs 32a, 32b and 32c apply resistance force to the nuclear fuel assembly at the same time, the hold-down margin is provided such that the gradient of the graph showing the spring force as a function of displacement is constant. Therefore, as shown in FIG. 3, to satisfy the hold-down margin under start-up conditions, the hold-down margin under hot full power conditions becomes excessively large. As a result of the excessive hold-down margin under hot full power conditions, the hold-down spring unit 32 applies an excessive resistance force to the nuclear fuel assembly, thus deteriorating the mechanical and structural stability of the nuclear fuel assembly. Accordingly, the present invention has been made keeping in mind the above mentioned problems occurring in the related art, and an object of the present invention is to provide a hold-down spring unit for a top nozzle of a nuclear fuel assembly which is configured such that the hold-down spring unit has the minimum hold-down margin under hot full power conditions, thus preventing the top nozzle from applying excessive resistance force to the nuclear fuel assembly, thereby enhancing the mechanical and structural stability of the nuclear fuel assembly, and a top nozzle for a nuclear fuel assembly having the hold-down spring unit. In order to accomplish the above mentioned object, the present invention provides a hold-down spring unit coupled to an upper end of a top nozzle of a nuclear fuel assembly, including: a first spring providing a hold-down force upon the nuclear fuel assembly under start-up conditions and hot full power conditions of a nuclear reactor; and a second spring providing an additional hold-down force upon the nuclear fuel assembly only under start-up conditions of the nuclear reactor. In order to accomplish the above mentioned object, the present invention provides a top nozzle for a nuclear fuel assembly, including: a coupling plate coupled to a guide thimble of the nuclear fuel assembly; a perimeter wall protruding upwards from a perimeter of the coupling plate, with a spring clamp provided on an upper surface of the perimeter wall; and a hold-down spring unit mounted to the upper surface of the perimeter wall. The hold-down spring unit includes: a first spring providing a hold-down force which acts upon the nuclear fuel assembly under start-up conditions and hot full power conditions of a nuclear reactor; and a second spring providing an additional hold-down force which acts upon the nuclear fuel assembly only under start-up conditions of the nuclear reactor. The first spring can have a first ramp, and a first neck part extending downwards from an upper end of the first ramp. The second spring can have a second ramp, a neck part insert slot formed through an upper end of the second ramp so that the first neck part of the first spring is inserted through the neck part insert slot, and a second neck part extending downwards from the upper end of the second ramp, the second neck part being shorter than the first neck part of the first spring. The hold-down spring unit can further include at least one intermediate spring provided under a lower surface of the first spring. The intermediate spring can have an intermediate ramp, and an intermediate neck part insert slot formed through an upper end of the intermediate ramp so that the first neck part of the first spring is inserted through the intermediate neck part insert slot. The first spring, the second spring and the intermediate spring can respectively include a first support part, a second support part and an intermediate support part. The first support part, the second support part and the intermediate support part can be respectively formed by bending the lower ends of the first ramp, the second ramp and the intermediate ramp in the horizontal direction. A first pin hole, a second pin hole and an intermediate pin hole can be respectively formed through the first support part, the second support part and the intermediate support part. The first spring, the intermediate spring and the second spring can have plate spring shapes. Hereinafter, an exemplary embodiment of the present invention will be described in detail with reference to the attached drawings. FIGS. 4A through 4C′ are exploded perspective views illustrating a hold-down spring unit 100 used in a top nozzle 30′ in FIG. 6. FIGS. 4A, 4B and 4C are bottom perspective views of a first spring 110, a second spring 130 and an intermediate spring 120; and FIGS. 4A′, 4B′ and 4C′ are top perspective views of the first spring 110, the second spring 130 and the intermediate spring 120. As shown in FIGS. 4A through 4C′, the hold-down spring unit 100 can include the first spring 110, the second spring 130 and the intermediate spring 120. The first spring 110 includes a first ramp 110′ which is a plate spring, a first support pant 111 which is provided on a lower end of the first ramp 110′, and a first neck part 112 which is bent downwards from an upper end of the first ramp 110′ and extends a predetermined length. A first pin hole 111a is formed through the first support part 111, so that a fastening pin 33 is inserted through the first pin hole 111a. A first neck part slot 113 is formed in the lower end of the first neck part 112 and extends a predetermined length in the vertical direction along the central axis of the first neck part 112. First neck part hooks 114 protrude outwards from the lower end of the first neck part 112 on opposite sides of the first neck part slot 113. The intermediate spring 120 includes an intermediate ramp 120′, an intermediate support part 121 which is provided on the lower end of the intermediate ramp 120′. A neck part insert slot 125 is formed through the upper end of the intermediate ramp 120′. An intermediate pin hole 121a is formed through the intermediate support part 121 so that the fastening pin 33 is inserted through the intermediate pin hole 121a. The second spring 130 includes a second ramp 130′ which is a plate spring, a second support part 131 which is provided on the lower end of the second ramp 130′, and a second neck part 132 which is bent downwards from the upper end of the second ramp 130′ and extends a predetermined length. The second neck part 132 is shorter than the first neck part 112. A second pin hole 131a is formed through the second support part 131 so that the fastening pin 33 is inserted through the second pin mole 131a. A neck part insert slot 135 is formed through the upper end of the second ramp 130′. Second neck part hooks 134 protrude outwards from opposite sides of the lower end of the second neck part 132. FIG. 5 is a partial side view of the top nozzle 30′, showing the assembled state of the hold-down spring unit 100 of FIG. 6. At the bottom of FIG. 5, the upper plate 40 is shown. At the left side of FIG. 5, the support end 101 is seen to hold together ends of the first spring 110 with its vertically oriented front neck part 112, the intermediate spring 120 and the second spring 130 with its vertically oriented second neck part 132. FIG. 6 is a perspective view of the top nozzle 30′ for the nuclear fuel assembly (hereinafter, referred to simply as ‘top nozzle’) according to the embodiment of the present invention. As shown in FIG. 6, the top nozzle 30′ includes a coupling plate 25, a perimeter wall 20 and a plurality of hold-down spring units 100. The coupling plate 25 is coupled to guide thimbles of the nuclear fuel assembly. The perimeter wall 20 protrudes upwards from the perimeter of the coupling plate 25. Spring clamps 31 and fastening parts (not shown) are provided on the upper surface of the perimeter wall 20. Spring insert holes 31a are formed in each spring clamp 31. The hold-down spring units 100 are provided on the upper surface of the perimeter wall 20 and the ends thereof are inserted into the corresponding spring insert holes 31a of the spring clamps 31. As shown in FIGS. 5 and 6, each hold-down spring unit 100 is configured such that the first spring 110, the intermediate spring 120 and the second spring 130 are assembled together in layers in such a way that the first neck part 112 is inserted through the neck part insert slot 125 of the intermediate spring 120 and the neck part insert slot 135 of the second spring 130. Here, the first support part 111 of FIG. 4A, the intermediate support part 121 of FIG. 4B and the second support part 131 of FIG. 4C form a support end 101 (refer to FIG. 5) having a layered structure. As shown in FIG. 6, to fasten the hold-down spring units 100, each of which has the layered structure attached to the top nozzle 30′, the support end 101 of each hold-down spring unit 100 is horizontally inserted into the corresponding spring hole 31a of the corresponding spring clamp 31. Thereafter, the fastening pin 33 is inserted into a corresponding fastening hole 33a of the spring clamp 31, the first pin hole 111a of FIG. 4A′, the intermediate pin hole 121a of FIG. 4B′ and the second pin hole 131a of FIG. 4C′ and fixed to the spring clamp 31 by a fixing method, such as welding, pining, nut or bolt coupling, etc., thus fastening the support end 101 (refer to FIG. 5) of the hold-down spring unit 100 to the top nozzle 30′ of FIG. 6. The lower end of the first neck part 112, which is inserted through the intermediate spring 120 and the second spring 130, is inserted and locked into an insert slot 41 of a corresponding upper plate 40 which is provided on the perimeter wall 20. The lower end of the second neck part 132 is also inserted and locked into the insert slot 41. Here, the first neck part slot 113 of FIG. 4A makes it possible to reduce the width of the lower end of the first neck part 112, thus facilitating the insertion of the first neck part 112 into the insert slot 41 of FIG. 6. The first neck part hooks 114 of FIG. 4A, which are provided on opposite sides of the first neck part 112, and the second neck part hooks 134 of FIG. 4C, which are provided on opposite sides of the second neck part 132, are inserted into the insert slot 41 of FIG. 6 and locked to the lower surface of the upper plate 40, thus preventing the first neck part 112 and the second neck part 132 from being undesirably removed from the insert slot 41. Separate locking pins may be locked to the first neck part hook 114 of FIG. 4A and the second neck part hook 134 of FIG. 4C so as more reliably to prevent the first neck part 112 of FIG. 4A and the second neck part 132 of FIG. 4C from being removed from the insert slot 41 of FIG. 6. In the hold-down spring unit 100 having the above-mentioned construction, the second neck part 134 of FIG. 4C is locked into the insert slot 41 of FIG. 6 and thus functions to prevent a loss of an element when the first spring 110 or the second spring 130 or the intermediate spring 120 is damaged. Furthermore, the hold-down spring unit 100 of the present invention functions to reduce hold-down margins within ranges of the positive values with respect to demand hold-down forces under hot full power conditions and start-up conditions when the nuclear reactor is operated. FIG. 7 is a graph showing the characteristic curve of the hold-down spring unit 100 according to the embodiment of the present invention. In the graph of FIG. 7, the dotted line shows the characteristic curve of the hold-down spring unit 32 of the conventional technique of FIG. 1. The solid line shows the characteristic curve of the hold-down spring unit 100 of FIGS. 4 through 6. As shown in FIG. 7, in the hold-down spring unit 100 according to the present invention, only the first spring 110 and the intermediate spring 120 are operated under the hot full power conditions, so that hold-down force exceeds the demand hold-down force but the hold-down margin (FPM) thereof is less than the hold-down margin (FPM′) of the conventional technique. Under the start-up conditions, after the first spring 110 and the intermediate spring 120 have entered the hold-down state, the second spring 130 also produces a hold-down force along with the first spring 110 and the intermediate spring 120. Thus, the hold-down margin is equal to or less than that of the conventional technique. Therefore, the hold-down spring unit 100 of the present invention minimizes the hold-down margin in a section of the hot full power conditions. This margin occupies most of the entire operating section of the nuclear reactor. In addition, under the start-up conditions, the hold-down spring unit 100 acts such that hold-down force having the hold-down margin equal to or less than that of the conventional technique is generated. Thus, optimal hold-down force can be applied to the nuclear fuel assembly. Thereby, when the nuclear reactor is in operation, excessive resistance force is prevented from being applied to the nuclear fuel assembly by the hold-down spring units 100. Hence, during the operation of the nuclear reactor, the top nozzle 30′ provides optimal hold-down force in response to any variation of the length of the nuclear fuel assembly, thus preventing the nuclear fuel assembly from being bent, and maintaining the position of the nuclear fuel assembly stably. As a result, the mechanical and structural stability of the nuclear fuel assembly can be enhanced. As described above, the present invention minimizes the hold-down margin under hot full power conditions of a nuclear reactor, thus preventing resistance force from being excessively applied to a nuclear fuel assembly from a top nozzle when the nuclear reactor is in operation. Furthermore, in the present invention, during the operation of the nuclear reactor, the top nozzle provides appropriate hold-down force in response to any variation of the length of the nuclear fuel assembly, thus maintaining the position of the nuclear fuel assembly more stably, and preventing the nuclear fuel assembly from being bent, thereby ensuring the mechanical and structural stability of the nuclear fuel assembly. Moreover, the present invention is configured such that, under start-up conditions of the nuclear reactor, the hold-down margin is equal to or less than the desired hold-down margin to provide sufficient hold-down force, and under hot full power conditions, the hold down margin is minimized. Thereby, the present invention prevents excessive hold-down force from being applied from the top nozzle to the nuclear fuel assembly when the nuclear reactor is in operation, thus enhancing the mechanical and structural stability of the nuclear fuel assembly in the entire operating section of the nuclear reactor. In addition, in the present invention, even if a hold-down spring unit is damaged, a loss of an element can be prevented by a neck part of a second spring. Although the preferred embodiment of the present invention has been disclosed for illustrative purposes, those skilled in the art will appreciate that various modifications, additions and substitutions are possible, without departing from the scope and spirit of the invention as disclosed in the accompanying claims.
049845101	claims	1. A system for posting articles into a containment maintained at sub-atmospheric pressure comprising a posting port in a wall of the containment having a sphincter seal and a removable lid, the sphincter seal being such as to engage articles posted through the port and to permit an inward air flow into the containment to oppose back-diffusion from the containment. 2. A system according to claim 1 in which the sphincter seal comprises an annular assembly of inner and outer brush seals between which are sandwiched rings of resilient material. 3. A system according to claim 2 in which each ring of resilient material is divided into sectors which extend radially inwardly beyound the inner radius of the brush seals. 4. A system according to claim 3 including a continuous elastic garter secured to the underside of the resilient material.
046631100	description	DESCRIPTION OF THE PREFERRED EMBODIMENTS A fusion reactor 10, illustrated in FIG. 1, includes a fusion chamber 12 and a blanket 14 permitting the breeding of fissile material which may be fabricated for use in a fission reactor. The blanket 14 includes a chamber wall 16 for isolating the fusion chamber 12 from the rest of the blanket 14, a neutron multiplication section 18, an enrichment section 20, and a reflector 22, in radially outward succession, respectively. The neutron multiplication section 18 includes a material which may produce about two neutrons upon bombardment by one neutron of high energy. The reflector 22 reflects neutrons back through the enrichment section 20 to increase the efficiency of the enrichment process. The enrichment section 20 includes fertile material, which is relatively dilute so as to limit fissioning and undue competition between fertile atoms for enriching neutrons. The enrichment section 20 also provides material capable of absorbing thermal neutrons in order to repress thermal fissioning of the bred fissile fuel. The amount of thermal neutron absorbing material is preferably selected to limit the thermal neutron flux without excessively competing with the fertile material for enriching neutrons. In accordance with the method of the present invention, fertile material is formed into particles, such as powder or pellets, and inserted into the enrichment section 20 of the fusion blanket 14. The particles may or may not be suspended in a slurry. The particles are exposed to neutrons produced directly and indirectly by the fusion process until the desired level of enrichment is achieved. The particles are then removed from the blanket 14, and any slurry carrier is removed. The particles may then be mixed to reduce or eliminate undesired nonuniformities in fissile enrichment. The fuel particles are then fabricated into appropriate form for use in fission reactors. The opportunities for diversion of weapons suitable fuel are minimized through such a method. Only fertile material would be fed into a fusion hybrid breeder and the bred fissile material would have a high content of fission products that would never be removed from the fuel, as it would never be reprocessed. These fission products would protect the bred fuel from diversion just as fission products currently protect spent fuel of a light water reactor (LWR) from diversion during shipping and storage. Should reprocessing be acceptable, the spent fuel element may be reprocessed to produce further fissile fuel for refabrication of a fuel element and/or further fertile material for further enrichment in the fusion blanket 14. Waste products of the reprocessing must then be properly disposed of. Describing the preferred fusion reactor 10 in greater detail, the fusion reactor 10 includes a chamber 12, which may be a plasma/vacuum chamber, wherein the fusion reaction takes place. The preferred reactor 10 is of the deuterium-tritium (D-T) fusion type. The fusion blanket 14 provides a chamber wall 16 for isolating the fusion chamber 12 from the remainder of the blanket 14. The blanket 14 includes four sections, a fusion chamber wall 16, a neutron multiplication section 18, an enrichment section, and a reflector 22. The chamber wall 16 serves to isolate the fusion chamber 12 from the remainder of the fusion blanket 14. The chamber wall 16 may also serve to contain the neutron multiplication material of the neutron multiplication region 18, where the material chosen achieves a liquid state during reactor operation. The preferred material for the chamber wall 16 is steel. In the illustrated embodiment, the chamber wall 16 has a diameter of about 4 m and is about 5 mm thick. Only one neutron is released per deuterium-tritium (D-T) reaction. The number of fusion neutrons would not suffice to breed fissile fuel efficiently, especially where it is expected that the fusion reactor 10 also breed enough tritium to be self-sustaining. The neutron multiplication section 18 serves to increase the number of neutrons available for breeding. By including a material which when bombarded by fusion neutrons produces two or more neutrons, there may be provided ample neutron bombardment for breeding both fissile and fusile materials. In addition, a fusion neutron has a very high energy of about 14.1 MeV, sufficient to induce direct fissioning of the fertile material. The direct fissioning of the fertile material spoils an enrichment opportunity and increases the actinide inventories of the eventual fuel particles. Accordingly, the neutron multiplication section 18 also serves to provide neutrons at more moderate energy levels so that direct fissioning is relatively infrequent. The multiplication section 18 may serve additional functions as well. A thermal neutron absorbing material may be included to reduce the number of thermal neutrons entering the enrichment section 20 and inducing thermal fissioning of bred fissile fuel. Tritium is bred by bombarding lithium with neutrons. Additionally, tritium may be bred in this section for recycling into the fusion reactor 10. Furthermore, heat generated in the multiplication section 18 must be transferred to cool the blanket 14 and provide energy to some destination. In the preferred embodiment, the multiplication section is primary of a lithium lead eutectic, such as Li.sub.17 Pb.sub.83. Structural components may be of ferritic steel. The Li.sub.17 Pb.sub.83 serves the functions described above. The lead, by way of Pb(n,2n) reactions serves as a neutron multiplier and a moderator of the very fast fusion neutrons. The lithium serves as tritium breeding material and as an absorber of thermal neutrons. Of course, the lithium may also moderate fast neutrons as well. Li.sub.17 Pb.sub.83 is a liquid at the operating temperatures of the fusion reactor 10 and thus serve as a coolant which is circulated to transfer heat from the blanket 14. In an embodiment in which the multiplier is solid, coolant pipes of ferritic steel could be incorporated as coolant passageways through which fluid coolant flows and transfer heat from the fusion blanket 14. Natural lithium is primarily .sup.7 Li, with about 7.4% .sup.6 Li. Inelastic scattering of .sup.7 Li by fusion neutrons, or other neutrons above 2.5 MeV, can result in their moderation and tritium production. However, the .sup.6 Li is the more active isotope in the context of the fusion reactor 10. The absorption of thermal neutrons by the .sup.6 Li isotope causes the isotope to break up into helium (.sup.4 He) with an energy of 2.0 MeV and tritons (.sup.3 H) with an energy of about 2.8 MeV. Upon capturing an electron, a triton becomes a tritium atom. In absorbing thermal neutrons, lithium also serves to limit the thermal flux in the adjacent enrichment section 20. The enrichment section 20 is designed to produce fissile atoms at an efficient enrichment rate, and to suppress thermal fissioning of the bred fissile atoms. The enrichment section 20 includes structural metal, material for moderating thermal neutrons-- and preferably for breeding tritium and fertile material. The fertile material may be .sup.232 Th and/or .sup.238 U. .sup.232 Th may absorb a neutron to produce .sup.233 U, a fissile material. .sup.238 U may absorb a neutron to produce .sup.239 Pu, also fissile. Various transuranic compounds may also be suitable fertile materials. The advantage of the thorium-uranium cycle is that .sup.232 Th is less prone to direct fissioning than .sup.238 U. Direct fissioning results in undesired actinide production while bypassing the production of a fissile atom. In the illustrated embodiment, .sup.232 Th in its oxide form, ThO.sub.2 , is employed as the fertile material. Preferably the ThO.sub.2 is in the form of ThO.sub.2 particles. The quantity of fertile material must be ample to support an appropriate ratio of fission to fusion reactors, and low enough so that competition among the fertile atoms does not unduly lengthen the time required for the target enrichment level. Since the enrichment rate is inversely proportional to the blanket 14 fertile inventory, the fertile residence time needed to achieve a given enrichment can be drastically reduced if the blanket fertile inventory is kept a minimum. A shorter residence time translates into a quicker fuel turnover, which is favorable from an economic standpoint. In the illustrated blanket 14, the enrichment section 20 includes between about 2% and about 3% ThO.sub.2. Preferably the concentration of ThO.sub.2 should be about 2.5%. The preferred concentration provides sufficient fuel to support about two HTGR's of power equal to that of the illustrated fusion reactor. If instead of ThO.sub.2, UO.sub.2 is used as the fertile material, we would need only about one-half or one-third as much fertile material in the blanket 14. The neutron capture cross section for uranium is about three times that of thorium. This results in about a two thirds reduction in residence time. Thus the quantity of UO.sub.2 may be between 0.5% and 1.5% of the enrichment section 20. Preferably, the quantity of UO.sub.2 is about 0.8% of the enrichment section 20. The enrichment section 20 contains thermal neutron moderating material, preferably one containing .sup.6 Li. The .sup.6 Li serves the same functions in the enrichment section 20 as it does in the multiplication section 18. However, the moderation of the thermal flux is more critical in the enrichment section 20 due to the proximity of the fertile particles. Conversely, the tritium breeding function is of lesser importance. The quantity of .sup.6 Li atoms should be adequate to moderate the thermal flux, and yet be slight enough not to compete unduly with the fertile particles for epithermal neutrons. Preferably the mean lithium concentration in the enrichment section 20 is on the order of 4.times.10.sup.20 atoms per cc. Note that fission suppression is achieved by: (1) the low level of fertile material in the blanket 14; (2) the presence of a neutron energy moderating material in the multiplication section 18, between the fusion chamber 12 and the fertile material to suppress fast fissioning of the fertile atoms; and (3) mixing the fertile particles with 1/v neutron absorbing materials, such as lithium compounds, to reduce the thermal fissioning of the accumulated bred fissile atoms. Among the advantages of this fission-suppressed aspect of the present invention are very low fission product and actinide inventories, and minimum thermal power swing during enrichment. The .sup.6 Li level in the lithium-lead compound is adjusted to about 1% of the enrichment section 20 to allow production of sufficient tritium, while not competing too severely with the .sup.232 Th for neutrons. The .sup.6 Li suppresses the thermal flux which would otherwise burn out the bred fuel. The .sup.6 Li may conveniently be included as a minority isotope in a lithium containing material. The concentration of .sup.6 Li may be determined by the amount of lithium containing material in the enrichment section 20 and by the density of lithium in the lithium containing material. Furthermore, the .sup.6 Li level in the lithium may be adjusted to effect the desired degree of suppression of the thermal flux. Where practical, it is preferable to adjust the .sup.6 Li level in the enrichment section 20 during the residence of the fertile material in the fusion reactor 10. By having negligible .sup.6 Li at the inception of the residence, a maximal number of neutrons is made available for enrichment. As fissile material is produced and the need for moderation of the thermal flux increases, .sup.6 Li may be added gradually. Thermal hydraulic or other considerations may dictate that the enrichment section 20 is not filled by fertile fuel, lithium material and structural metal. Accordingly, it may be advantageous to include other material, such as graphite and/or silicon carbide in the enrichment section 20. Means for transferring heat from the enrichment section 20 must also be provided. The particles of fertile fuel may be packed in breeding chambers in the enrichment section 20 or suspended in a slurry circulated in pipes extending through the enrichment section 20. In the first, or "dry" embodiment, breeding chambers may extend vertically from the top to adjacent the bottom of the radial portion of the fusion blanket 14. The breeding chambers are formed in solid material of the enrichment section 20 which may include graphite and/or silicon carbide. The breeding chambers may extend vertically through the enrichment section 20, and may have diameters about 2 cm. Preferably, the enrichment section 20 includes a lithium material, such as Li.sub.7 Pb.sub.2, LiAlO.sub.2, Li.sub.2 O, and mixtures thereof, which remains solid at fusion blanket operating temperatures. Li.sub.7 Pb.sub.2 is preferred because the lead serves to supplement the neutron multiplication collected in the multiplication section 18. Lithium aluminate, LiAlO.sub.2, is advantageous because of its high melting point. Lithium oxide, Li.sub.2 O, is advantageous because it is a dense lithium containing compound. LiO.sub.2 might be used in designs where it is desired to maximize the amount of graphite or silicon carbide in the enrichment section 20. More graphite or silicon carbide might be required by neutron moderation and thermal hydraulic or heat transfer considerations. To recover bred tritium in the first preferred embodiment, fluid, such as helium, may flow through pellets of the lithium containing material to collect the tritium. This process may be facilitated by forming holes in the pellets through which the collector fluid flows. The collector fluid preferably flows more slowly than the coolant fluid. In the first embodiment, the fuel particles may be enriched to the level required for operation of an HTGR, which may be about 4%. The duration of the enrichment process is a function of the power output of the fusion reactor 10. If the illustrated fusion reactor 10 produces a 4 MW/m.sup.2 neutron loading at the chamber wall 16, known methods of calculation suggest that about 2 months of residence time will be required for each percent of enrichment. Accordingly, about eight months are required to achieve the target 4% enrichment level. After enrichment, the fusion reactor 10 may be shut down and the fuel particles removed. The removed particles may be mixed to compensate for nonuniformities in enrichment introduced as a function of location of particles within the enrichment section 20. The mixing may be random or systematic, where the enrichment distribution is established. The removal of fuel particles from the fusion blanket 14 may be effected by blowing the particles out by gas injection or by mechanical means. Alternatively, the fuel particles may be included within a liner which is inserted into the breeding chambers of the blanket 14. After enrichment, the liner may be removed along with the contained fuel particles. The liner may be of steel, such as stainless steel. Where the enrichment profile is known, the fuel particles may be collected in two or more aggregates with different average enrichments. The particles within each aggregate may then be mixed so that the enrichment is uniform within each aggregate. This technique may be used where different enrichment levels are required for different fission reactors, or different components in a single fission reactor. In some HTGRs, for example, the fuel in the control rods must be more highly enriched than the fuel in the fuel blocks. In these cases, the fuel for the control rods may be collected from a relatively high enrichment region of the enrichment section 20 and the fuel for the fuel blocks from a relatively low enrichment region of the enrichment section 20. The fuel collected from the relatively high enrichment region may be mixed and then imbedded into graphite matrices to fabricate fuel rods. The fuel collected from the relatively low enrichment region may be randomly mixed and inserted into the breeding chambers of the fuel blocks. In the second preferred or "wet" embodiment, the fertile particles may be suspended in a carrier fluid such as Li.sub.17 Pb.sub.83 eutectic, which has a melting point of about 235.degree. C. The Li.sub.17 Pb.sub.83 may also serve as the coolant for the enrichment section 20. The Li.sub.17 Pb.sub.83 is a preferred carrier because it has a density comparable to that of ThO.sub.2, so that the particles are more evenly distributed throughout the slurry. The .sup.6 Li in the slurry helps moderate the thermal flux and suppress thermal fission. The advantage of having a liquid carrier for the particles is that the void spaces are filled between the particles and where the liquid is a conductor, like Li.sub.17 Pb.sub.83, heat transfer to the walls of pipes carrying the slurry is facilitated. Thus hot spots may be alleviated and the maximum reactor temperatures lowered to reduce materials stresses and other reactor design problems. Li.sub.17 Pb.sub.83 may be circulated out of the normal coolant passageways, to allow recovery of the bred tritium. The amount diverted to recovery may be about 1% to 2% of the total flow. The diversion permits a small quantity of fluid to flow slowly enough for the recovery process to be effectuated. The amount of coolant is determined by thermal hyrdraulic considerations. Outside the pipes, the bulk of the enrichment section 20 may also be Li.sub.17 Pb.sub.83. The enrichment section may also include solid moderating material such as silicon carbide or graphite. Silicon carbide is preferred over graphite, which reacts with Li.sub.17 Pb.sub.83 at temperatures above 500.degree. C., whereas silicon carbide is relatively inert below 800.degree. C. In the second preferred embodiment, the .sup.6 Li level may be adjusted as the Li.sub.17 Pb.sub.83 circulates. Thus, the .sup.6 Li concentration may be negligible at the onset of enrichment so as not to compete with the fertile material for thermal neutrons. As the fissile material content increases as a function of the enrichment of the fertile material, more .sup.6 Li lithium may be added gradually to suppress thermal fissioning of the fissile fuel. In the second embodiment, the fuel particles may be enriched to a level suited to a LWR, such as a 3.3% enrichment level, equal to 3% enrichment plus another 10% to compensate for the effect of poisons in the LWR. Accordingly, a residence time of about 6.6 months would be appropriate in the second preferred embodiment. Once the desired enrichment level is achieved, the slurry may be removed from the blanket pipes. The bulk of the carrier may be strained away. Nitric acid may be used to clean the particles of the remaining eutectic. The particles may then be mixed, randomly or systematically, to achieve uniformity. The particles may then be fabricated into fuel elements such as fuel pins for a LWR. While the first embodiment is described as producing fuel for an HTGR and the second embodiment is described as producing fuel for a LWR, either embodiment could be adapted for breeding fuel for a variety of fission reactors. The embodiments may have similar chamber walls, multiplication sections and reflectors. The preferred dimensions, common to both embodiments are: the diameter of the chamber wall 16 is about 4 m; the thickness of the chamber wall 16 is about 5 mm; the thickness of the multiplication section 18 is about 0.4 m; the thickness of the enrichment section 20 is about 0.3 m; and the thickness of the reflector 22 is about 0.3 m. The neutronics performance of the wet and dry embodiments with the dimensions as described above is summarized in Table I. The results of the neutronics calculations are given in this table for the blankets when the .sup.233 U enrichments are 0%, 2% and 4%. TABLE I ______________________________________ SUMMARY OF NEUTRONICS PERFORMANCE .sup.233 U Enrichment Beginning End of of Life Life 0% 2% 4% ______________________________________ .sup.6 Li(n,.alpha.) T (T.sup.6) multiplication section 0.6342 0.6474 0.6608 enrichment section 0.4076 0.4181 0.4304 Total 1.0418 1.0655 1.0912 .sup.7 Li(n,n'.alpha.) T (T.sub.7) multiplication section 0.02416 0.02418 0.02419 enrichment section 0.00316 0.00370 0.00428 Total 0.02732 0.02788 0.02847 Total T.sub.6 + T.sub.7 1.0691 1.0934 1.1197 .sup.233 U Production (U/n) 0.4082 0.4221 0.4365 .sup.233 U(n,f) (fission/n) 0.0 0.0459 0.0953 Net .sup.233 U Production 0.4082 0.3762 0.3412 Blanket Energy 1.20 1.78 2.42 Multiplication Other Reactions (R/n) Pb(n,2n) 0.6751 0.6752 0.6752 Fe(n,.gamma.) 0.1140 0.1170 0.1202 Th(n,2n) 4.15 .times. 4.17 .times. 10.sup.-4 4.20 .times. 10.sup.-4 10.sup.-4 Th(n,3n) 8.79 .times. 8.61 .times. 10.sup.-5 8.45 .times. 10.sup.-5 10.sup.-5 Th(n,f) 2.27 .times. 3.20 .times. 10.sup.-4 4.17 .times. 10.sup.-4 10.sup.-4 .sup.233 U(n,2n) 0.0 2.77 .times. 10.sup.-6 5.79 .times. 10.sup.-6 .sup.233 U(n,3n) 0.0 1.21 .times. 10.sup.-7 2.42 .times. 10.sup.-7 .sup.233 U(n,.gamma.) 0.0 6.51 .times. 10.sup.-3 0.0134 ______________________________________ As shown in Table I, the net .sup.233 U production rate is 0.4082 when the blanket 14 is fresh, and drops to 0.3762 and 0.3412 when the blanket 14 is enriched with 2% and 4% .sup.233 U, respectively. The reduction of the .sup.233 U production rate at the end of the blanket exposure, i.e., 4% .sup.233 U enrichment, is about 16%. The tritium breeding ratio increases from 1.0691 to 1.1197 when the blanket 14 is at fresh and 4% .sup.233 U enrichment, respectively. This demonstrates that adequate tritium regeneration, an average of about 1.1 tritons per D-T neutron, can be obtained by the blanket design. The blanket energy multiplication, which is the ratio of the blanket total nuclear heating to the fusion neutron energy, is 1.20 at the beginning of blanket life. It increases to 1.78 and 2.42 when the .sup.233 U enrichment in the blanket 14 reaches 2% and 4%, respectively. The increase of the blanket energy multiplication is due solely to the fissioning induced in the accumulated .sup.233 U in the blanket 14. The increase at 4% .sup.233 U enrichment from the multiplication at the beginning of blanket life is about 100%. However, this thermal heat swing is modest compared to that of a fast-fission hybrid blanket, and is manageable without much effort in the blanket thermal design. Also shown in Table I are the reaction rates for several important nuclear reactions such as Pb(n,2n), Th(n,2n), and Th(n,3n). The Th(n,2n), Th(n,3n) and Th(n,f) reaction rates, 4.15.times.10.sup.-4, 8.79.times.10.sup.-5 and 2.27.times.10.sup.-4, respectively, per D-T neutron, are very small. The Th(n,f) reaction rate increases with the increase of .sup.233 U enrichment in thorium, and is about twice that at fresh, when the blanket 14 reaches 4% .sup.233 U enrichment. However, the contribution of the Th(n,f) reaction to the blanket energy multiplication is so small, less than 0.2%, that it can be ignored. These low actinide production and thorium fission rates are basically due to the dilution of the thorium inventory and the suppression of the high-energy neutrons in the blanket design. However, because of the reduction of the thorium inventory, the relative concentrations of the actinide elements in the thorium or bred uranium are not so significantly reduced. The .sup.232 U concentration in the bred .sup.233 U at 4% .sup.233 U enrichment is estimated to be about 130 ppm, which is about two orders of magnitude lower than in a typical fast-fission blanket. The concentrations of other possible uranium isotopes, such as .sup.234 U, are much lower than the .sup.232 U concentration. Hence, production of very high quality fissile fuel can be expected in the illustrated blanket 14. The average .sup.233 U production in the illustrated blanket 14 is about 0.38 U/n. FIG. 2 depicts the blanket exposures (MW-yr/m.sup.2) required to reach 4% .sup.233 U enrichment in thorium as a function of the .sup.233 U production rate, and may be used to calculate the necessary residence times for fertile fuel. Given a neutron wall loading of about 4.0 MW/m.sup.2, the residence time of the blanket 14 in the reactor 10 is about eight months, which is significantly shorter than the residence times required in most hybrid blanket designs. Without reprocessing, one fusion reactor can support about two fission reactors of power equal to that of the fusion reactor. With reprocessing, ten to twenty fission reactors may be supported, depending upon design. The higher numbers may be obtained where the fission reactor is itself a breeder. Should reprocessing be acceptable, the spent fuel element may be reprocessed to produce further fissile fuel for refabrication of a fuel element and/or further fertile material for further enrichment in the fusion blanket 14. Waste products of the reprocessing must then be properly disposed of. In accordance with the above, an improved method of breeding fissile fuel and a fusion blanket for practicing the method are provided. Among the advantages of the present invention are the following. No reprocessing is needed, thus easing the potential political complications of the fusion-fission hybrid reactor development, and the fuel production cost will be lower than when reprocessing is required. Reprocessing of the spent fuel may be added at a later date. A relatively short fertile residence time is required, so that the probability of blanket failure is reduced, materials requirements for the blanket are relaxed, and fuel turnover time is lessened. The thermal power swing is maintained within a factor of about two, well within the capabilities of the materials incorporated in the inventive blanket design. The fission-suppressed blanket design is used to maintain a high enriching rate, low fission product and actinide inventories, and modest thermal power swing during enrichment. Tritium is produced for refueling the fusion reactor 10. The above and other embodiments are within the spirit and scope of the present invention.
	description	This application claims priority under 35 U.S.C. §119 to German Patent Application No. DE 10 2007 058 777.7 filed in Germany on Dec. 6, 2007, the entire content of which is hereby incorporated by reference in its entirety. A method is disclosed for commissioning pneumatically operated actuators that are controlled by a positioner. Linear drives and rotary drives are used in automation engineering and differ from each other in the way the final control element is actuated. These different types of drive require different forms of control by the positioner. To achieve this, a parameter is entered manually during commissioning of the pneumatically operated actuator that specifies whether a linear drive or a rotary drive is connected to the positioner. This procedure is prone to errors, and if an incorrect entry is made can result in damage to the actuator and/or the final control element. Exemplary embodiments disclosed herein can improve the commissioning of the known pneumatically operated actuator by detecting the drive type automatically. A method is disclosed for commissioning pneumatically operated actuators that are controlled by a positioner, wherein a constant flow of pneumatic fluid is applied to the actuator during commissioning, while the pneumatic fluid is applied, a drive-specific characteristic curve of the fed back position is recorded over time, the measured characteristic curve is compared with a given specimen characteristic curve, and the drive type of the actuator is inferred from the level of difference or agreement between the drive-specific characteristic curve and the specimen characteristic curve. In another aspect, an arrangement is disclosed for commissioning an actuator. Such an arrangement comprises: a positioner capable of determining a drive type based on a constant flow of pneumatic fluid applied to the actuator; a lifting rod mechanically caused to be moved by the constant flow of pneumatic fluid to the actuator; and a position sensor that senses the movement of the lifting rod for feedback signaling to the positioner. The positioner records data for a drive-specific characteristic curve based on the feedback signal for comparison of the measured characteristic curve with a given specimen characteristic curve to infer the drive type. The disclosure is based on a pneumatically operated actuator, which is connected to a positioner and controlled by this positioner, with the position of the drive of the actuator being fed back to the positioner. According to an exemplary embodiment of the disclosure, to determine the drive type, a constant flow of pneumatic fluid is applied to the actuator during commissioning while a drive-specific characteristic curve of the fed back position is recorded over time. Then the measured characteristic curve is compared with a given specimen characteristic curve. The drive type is inferred from the level of difference or agreement between the drive-specific characteristic curve and the specimen characteristic curve. The shape of the characteristic curve of a rotary drive differs significantly from the shape of the characteristic curve of a linear drive. The differences are easily exposed by comparing with a given specimen characteristic curve. According to another exemplary embodiment of the disclosure, it is provided that the given specimen characteristic curve is determined by idealizing known characteristic curves of one of the two drive types to be distinguished. Where the observed drive type matches the drive type on which the specimen characteristic curve is based, the level of agreement between the characteristic curves is significantly high. Where the observed drive type differs from the drive type on which the specimen characteristic curve is based, its characteristic curve differs significantly from the specimen characteristic curve. According to yet another exemplary embodiment of the disclosure, it is provided that the level of agreement between the characteristic curve of the observed actuator and the specimen characteristic curve is determined by cross-correlation. As shown in FIG. 1, a process valve 2 is fitted in a pipeline 1, a section of which is shown, of a process engineering plant, which is not shown further. Inside the process valve 2 is a closing body 4 that interacts with a valve seating 3 to control the amount of process medium 5 that passes through. The closing body 4 is operated linearly by an actuator 6 via a lifting rod 7. The actuator 6 is connected to the process valve 2 via a yoke 8. A positioner 9 is mounted on the yoke 8. The travel of the lifting rod 7 is signaled to the positioner 9 via a position sensor 10. The detected travel is compared in a control unit 18 with the setpoint value supplied via a communications interface 11, and the actuator 6 is controlled as a function of the determined control error. The control unit 18 of the positioner 9 comprises an I/P converter for converting an electrical control error into an appropriate control pressure. The I/P converter of the control unit 18 is connected to the actuator 6 via a pneumatic fluid supply line 19. During commissioning, a constant flow of pneumatic fluid is applied to the actuator 6 by the positioner 9 in order to determine the drive type. This causes the lifting rod 7 to move, and this movement is signalled to the positioner 9 by the position sensor 10. In the positioner 9, a drive-specific characteristic curve of the fed back position of the lifting rod 7 is recorded over time. The recorded characteristic curve is compared with a given specimen characteristic curve. The drive type is inferred from the level of difference or agreement between the drive-specific characteristic curve and the specimen characteristic curve.yl=xl(l) for l=0 . . . (n−1) The y-values are compared with an ideal function for a linear drive. The cross-correlation is defined by the coefficients C l = ∑ k = 0 n - 1 ⁢ y k ⁢ Z k + l In addition, assembly errors are detected by comparing the characteristic curve of the observed actuator with the specimen characteristic curve using cross-correlation, these errors being revealed by a shift in the characteristic curve by a fixed amount (offset) compared with the specimen characteristic curve. These errors are advantageously detected using the same means as those provided for detecting the drive type. The result of the absolute position measurement is thereby improved. It will be appreciated by those skilled in the art that the present invention can be embodied in other specific forms without departing from the spirit or essential characteristics thereof. The presently disclosed embodiments are therefore considered in all respects to be illustrative and not restricted. The scope of the invention is indicated by the appended claims rather than the foregoing description and all changes that come within the meaning and range and equivalence thereof are intended to be embraced therein. 1 pipeline 2 process valve 3 valve seating 4 closing body 5 process medium 6 actuator 7 valve rod 8 yoke 9 positioner 10 position sensor 11 communications interface 18 control unit 19 pneumatic fluid supply line
	abstract	A thermonuclear reactor is provided having a vacuum casing and blanket modules connected thereto with flexible supports. The flexible supports are formed from a material with high electrical conductivity. Each flexible support is secured at one end on the vacuum casing and at the other end on a blanket module, the two secured ends of each flexible support face the blanket module. The flexible support is formed from two hollow cylindrical elements placed one in the other and perforated by longitudinal slots in a part free from mountings. The ends of the hollow cylindrical elements opposite the secured ends are connected electrically and mechanically. The technical result consists in diverting eddy currents away from a blanket module of a thermonuclear reactor and simultaneously eliminating electrical connectors from the composition of a blanket and reducing bunching on a blanket module side facing the vacuum casing.
043354670	claims	1. A liquid metal cooled nuclear reactor comprising in combination (a) a vessel for receiving the reactor core and said liquid metal; (b) at least one heat exchanger for ensuring a heat exchange between said liquid metal and a second fluid; (c) a first duct for connecting said vessel with the inlet of the heat exchanger, said duct opening into said vessel above the core; and (d) a second duct for connecting said vessel with the outlet of said heat exchanger said vessel being provided, along the outer periphery thereof, with a first supporting means situated between said ducts said first duct being substantially rectilinear and horizontal and opening into said vessel above, and in the vicinity of, said first supporting means, said heat exchanger provided with a second supporting means at a level that is above the reactor support level wherein the stresses resulting from differential expansions of the ducts are reduced and constituted by at least two supporting members diametrically opposed with respect to the vertical axis of the exchanger outer cover, each of said supporting members comprising a horizontal fixed support plate, a horizontal backing plate integral with said exchanger and situated above said support plate, and a first plurality of rollers between said plates, the axes of revolution of said rollers being at right angles to the direction of said tubing; and wherein each supporting member comprises a second series of rollers parallel to said first series, said second rollers being arranged above said backing plate and bound to said supporting plate so mounted that there is a clearance between said rollers and said backing plate such supporting member also comprising a horizontal upper plate for the abutment of the rollers of said second series, said plate being integral with said support plate and so arranged that whenever the heat exchanger assumes an abnormal movement, said rollers of the second series come into abutment with said horizontal upper plate as well as a third series of rollers with vertical axis with respect to the reactor, adapted to form an abutment, in the case of horizontal movements, between said backing plate and a vertical extension of said supporting plate. 2. A nuclear reactor as in claim 1, wherein said vessel comprises a main outer vessel provided with said first supporting means, a primary inner vessel coaxial with said outer vessel, said primary vessel containing said core and provided with third supporting means constituted by first parts integral with the main vessel inner surface and by second parts integral with the primary vessel outer surface and resting on said first parts, said third supporting means being situated at said predetermined level, wherein said first duct opens into said primary vessel above said core and passes through the annular space between said main vessel and said primary vessel, and wherein said second duct opens into said annular space situated at a lower level below the support means. 3. A nuclear reactor as in claim 1, wherein the centers of gravity of both the heat exchanger and the reactor vessel are situated below the respective supporting levels.
	abstract	A method of loading fuel in multiple reactor cores associated with a plurality of fuel cycles. The method includes, in a first fuel cycle, loading a first reactor core with a first fuel assembly selected from a first batch of fuel, loading the first reactor core with a first partially spent fuel assembly from a second batch of fuel, loading a second reactor core with a second fuel assembly from the first batch of fuel, and loading the second reactor core with a second partially spent fuel assembly from the second batch of fuel. In a second fuel cycle, which is performed after a completion of the first fuel cycle, the method includes loading the second reactor core with a fresh fuel assembly, and loading the second reactor core with the first fuel assembly from the first batch of fuel.
	summary
	abstract	An X-ray condenser for condensing X-rays radiated from an X-ray source to a very small condensing spot is disclosed. X-rays from the X-ray source are formed to a parallel X-ray beam by a parallel type parabolic reflection mirror. The parallel X-ray beam is made monochromatic by an analyzing crystal and condensed to the condensing spot by a zone plate. The zone plate is constructed by alternately arranging a plurality of X-ray transmitting bands and a plurality of X-ray shielding bands and can condense the parallel X-ray beam to a very small focus point.
	claims	1. A control rod/fuel support handling apparatus that removes and attaches a control rod disposed between fuel assemblies in a reactor pressure vessel and a fuel support installed on a core plate located below a upper guide in the reactor pressure vessel, the control rod having a lower end part detachably connected to a control rod drive mechanism via a bayonet coupling, the fuel support supporting a lower end part of the fuel assembly and allowing insertion of the control rod in a cruciform space,the control rod/fuel support handling apparatus comprising:a fuel support gripper that is supported from above of the reactor pressure vessel so as to move in a vertical direction and holds the fuel support with a support gripping member so as to remove and attach the fuel support;a control rod gripper that is accommodated in the fuel support gripper, includes a vertically movable and rotatable elevating/rotating unit, holds the control rod with a control rod gripping member provided under the elevating/rotating unit, disconnects and connects the control rod and the control rod drive mechanism, and allows removal and attachment of the control rod through the cruciform space of the fuel support;a fuel support seating detector which detects seating of the lower end of the support gripping member on the core plate;a fuel support grip detector which detects that the support gripping member holds the fuel support;a load measuring instrument that is provided in the elevating/rotating unit and measures a load applied to the control rod gripping member;an operation mechanism that provides instructions on operations of the fuel support gripper and the control rod gripper, the operation mechanism configured to be operated by a first signal from said fuel support seating detector, a second signal from said fuel support grip detector, and a third signal from said load measuring instrument; anda control panel that controls the fuel support gripper in response to the instructions from the operation mechanism and automatically controls the control rod gripper. 2. The control rod/fuel support handling apparatus according to claim 1, wherein the fuel support gripper comprises a gripper frame having a wire connecting part connecting a suspension wire on an upper end of the gripper frame, the support gripping member is provided so as to open and close on a lower end of the gripper frame, and the gripper frame accommodates and supports the control rod gripper so as to move in the vertical direction. 3. The control rod/fuel support handling apparatus according to claim 2, wherein the fuel support gripper comprises:a fuel support seating detector that detects that the lower end of the gripper frame has been seated on the core plate; anda fuel support grip detector that detects that the fuel support has been held by the support gripping member. 4. The control rod/fuel support handling apparatus according to claim 2, wherein a gripper body of the control rod gripper or the gripper frame of the fuel support gripper comprises an upper limit position detector that detects that the gripper body is supported by the gripper frame and the gripper frame has reached an upper limit position with respect to the gripper body. 5. The control rod/fuel support handling apparatus according to claim 1, wherein the control rod gripper comprises:a gripper body;an elevating/rotating unit provided with the control rod gripping member opened and closed on a lower end of the elevating/rotating unit;a lift cylinder that lifts and lowers the elevating/rotating unit with respect to the gripper body;a rotating mechanism that rotates the elevating/rotating unit a predetermined angle around an axis;a lifting/lowering cylinder that is provided on an upper end of the gripper body, has a piston end fixed on a locking plate lockable to the upper guide, and allows lifting and lowering of the gripper body; anda load measuring instrument that is provided in the elevating/rotating unit and measures a load applied to the control rod gripping member. 6. The control rod/fuel support handling apparatus according to claim 5, wherein the gripper body of the control rod gripper or the gripper frame of the fuel support gripper comprises an upper limit position detector that detects that the gripper body is supported by the gripper frame and the gripper frame has reached an upper limit position with respect to the gripper body. 7. The control rod/fuel support handling apparatus according to claim 5, wherein the control rod gripper comprises:an ascent/descent position detector that detects an ascent/descent position of the elevating/rotating unit lifted or lowered by the lift cylinder;a rotational position detector that detects a rotational position of the elevating/rotating unit rotated by the rotating mechanism;a lifted/lowered position detector that detects a lifted/lowered position of the gripper body lifted or lowered by the lifting/lowering cylinder;a control rod seating detector that detects that the lower end of the elevating/rotating unit has been seated on the control rod; anda control rod grip detector that detects that the control rod has been held by the control rod gripping member,wherein the control panel automatically controls the control rod gripper based on detection data of these detectors. 8. The control rod/fuel support handling apparatus according to claim 7, wherein the ascent/descent position detector comprises:an ascent position detector that detects the ascent position of the elevating/rotating unit; anda descent position detector that detects the descent position of the elevating/rotating unit. 9. The control rod/fuel support handling apparatus according to claim 7, wherein the rotational position detector comprises:a counterclockwise limit position detector that detects a counterclockwise rotation limit position of the elevating/rotating unit;a clockwise limit position detector that detects a clockwise rotation limit position of the elevating/rotating unit;an intermediate first position detector that detects that the elevating/rotating unit has reached an intermediate first position between the limit positions; andan intermediate second position that detects that the elevating/rotating unit has reached an intermediate second position between the limit positions. 10. The control rod/fuel support handling apparatus according to claim 7, wherein the lifted/lowered position detector comprises:a lifted position detector that detects the lifted position of the gripper body; anda lowered position detector that detects the lowered position of the gripper body.
	description	1. Field of the Invention The present invention relates to a radiation monitor and particularly to a radiation monitor which confirms soundness of a steam generator in a pressurized water reactor plant. 2. Description of the Related Art There is a radiation monitor which is called a sensitive main steam pipe monitor among radiation monitors which confirm soundness of a steam generator by monitoring leakage from a primary coolant to a secondary coolant of a steam generator (SG) of a nuclear power plant. This sensitive main steam pipe monitor includes: a radiation detector which is disposed close to a main steam pipe and detects a radiation to output an analog voltage pulse; and a count-rate measurement unit which receives the analog voltage pulse, discriminates the analog voltage pulse entering a high-energy window which is set to contain a photoelectric peak, a single-escape peak, and a double-escape peak of γ-ray (6.13 MeV) of N-16 which is a radionuclide contained in the steam in the main steam pipe, to output a digital pulse, and measures a count rate of the digital pulse, and monitors a change in the count rate. In the same manner as a typical count-rate measurement unit of a radiation monitor, the count-rate measurement unit of the sensitive main steam pipe monitor counts digital pulses which are discriminated by pulse heights, and acquires and outputs a count rate by performing a time constant process using software so that a standard deviation becomes constant based on the counted value. It is also possible to have a suitable response according to the purpose, by switching the standard deviation according to the count rate. If necessary, the plurality of count rates can be acquired by performing the plurality of time constant processes, and the plurality of count rates having different standard deviations can be displayed for comparison (for example, see PTL 1). In addition, a technology of determining whether an indication increase occurs by synchronizing the upper stream and the lower stream of the main steam pipe in two detecting positions, depending on a signal or noise has also been proposed (for example, see PTL 2). The steam in the main steam pipe is in a secondary system and does not contain artificial radionuclides in a normal state. In addition, a background count rate in a normal state is low as approximately several cpm because cosmic radiations are dominant, and the background count rate and an alert setting point are close to each other. Accordingly, when alert transmission is attempted at high precision by preventing erroneous alerts, the standard deviation is reduced, and as a result, the response of the alert transmission is delayed, and when the standard deviation is increased by giving priority to the response of the alert transmission, erroneous alerts may frequently occur. Therefore, the alert is divided into two stages which are a caution alert and a high-level alert which is at an upper level of the caution level. The caution alert is transmitted during a stage of a slight leakage, and investigation is minutely performed by including a possibility of the erroneous alerts. [PTL 1] JP-A-61-128184 {Expression (1), Expression (4), FIGS. 1, 5, 6, and 10 to 16} [PTL 2] JP-A-4-268496 (FIGS. 1 and 2) The radiation monitor of the related art is configured as described above. Since the analog voltage pulse from the radiation detector is input to the count-rate measurement unit, the pulse height values entering the set window are discriminated and counted, the time constant process is performed us ing software so that the standard deviation becomes constant based on the counted value, the count rate is acquired and output by giving priority to the responsiveness, and the alert setting point is close to the background count rate, the alert may be erroneously transmitted due to a statistical change, so-called fluctuation, of the count rate, and it is necessary to perform an operation of performing off-line inspection of an apparatus, to be safe, to confirm soundness, even when the count rate is restored to the background count rate. With respect to this, in a method of acquiring the plurality of count rates having different standard deviations from the same input, for comparing the change thereof, approximately 20 minutes are taken for the regular output of the count rate obtained by giving priority to the responsiveness to approach an apex of the fluctuation, and when this changes in a state where the count rate for diagnosis having low responsiveness follows, it is difficult to identify the reason because the input is in the same pulse stream. In addition, as disclosed in PTL 2, in the proposal of comparing changes of count rates of two detecting positions of the upper stream and the lower stream of the main steam pipe, the background count rate is small as a several cpm, and accordingly, a possibility that increasing tendencies become the same, cannot be ignored, and there is no fundamental resolution disclosed. The present invention has been made to address the aforementioned problems and to provide a radiation monitor having high reliability and excellent maintainability which accurately determines whether or not fluctuation is a reason by online self-diagnosis with respect to transmission of a caution alert and provides information regarding the result. According to the invention, there is provided a radiation monitor including: a radiation detector which detects a γ ray emitted from a measurement target nuclide and outputs an analog voltage pulse; and a radiation measuring instrument which receives the analog voltage pulse output from the radiation detector, and measures and outputs radiation in a measurement energy range, in which the radiation measuring instrument includes a pulse amplifier which amplifies the input analog voltage pulse and removes superimposed high frequency noise, a high-energy count-rate measuring instrument which discriminates the analog voltage pulse output from the pulse amplifier by a high-energy window and a low-energy window which are set so as not to be superimposed on each other in accordance with a voltage level, respectively, measures and outputs a high-energy count rate by performing a time constant process of the pulses entering the high-energy window so that a standard deviation becomes constant, and outputs an alert, when the high-energy count rate is increased beyond an acceptable set value, a low-energy count-rate measuring instrument which measures and outputs a low-energy count rate by moving and averaging the pulse entering the low-energy window at a constant measurement time, an alert-diagnosis device which determines whether or not the low-energy count rate is in a set acceptable range, when an alert is output from the high-energy count-rate-measuring instrument, determines that the alert is caused by fluctuation, when the low-energy count rate is in the acceptable range, determines that the alert is caused by any one of an increase in the γ ray which is a measurement target or enter of noise, when the low-energy count rate increases beyond the acceptable range, and outputs a result of the determination, and a display/user-operation device which displays each output and performs operations and settings of each unit. The radiation monitor according to the invention is provided to automatically determine and display whether the alert is caused by the fluctuation or other matters, and therefore, a radiation monitor having high reliability and maintainability in which the time necessary for investigation of the causes of the alert transmission is significantly shortened, is obtained. The foregoing and other objects, 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. Hereinafter, preferred embodiments of the radiation monitor according to the invention will be described with reference to the drawings. FIG. 1 is a diagram showing a configuration of a radiation monitor according to Embodiment 1 of the invention. In FIG. 1, a radiation detector 1, which is radiation detecting means, detects a γ ray emitted from N-16 nuclide which is a measurement target nuclide and outputs an analog voltage pulse. A radiation measurement unit 2, which is radiation measurement means, includes a pulse amplifier 21 which is pulse amplification means, a high-energy count-rate-measurement functional unit 22a which is high-energy count-rate-measurement means, a low-energy count-rate-measurement functional unit 23 which is low-energy count-rate-measurement means, an alert-diagnosis functional unit 24 which is alert-diagnosis means, an interface functional unit 25, and a display/user-operation put unit 26 which is display/user-operation means. The pulse amplifier 21 receives and amplifies the analog voltage pulse output from the radiation detector 1 and removes superimposed high-frequency noise and outputs the pulse. The high-energy count-rate-measurement functional unit 22a includes a high-window pulse-height discriminator 221, a high counter 222, and a high-energy count-rate-operation functional unit 223a, the high-window pulse-height discriminator 221 receives the analog voltage pulse output from the pulse amplifier 21 and discriminates the pulse entering the window having the set high energy to output a digital pulse, and the high counter 222 counts the digital pulse at fixed cycle and outputs a counted value. In addition, the high-energy count-rate-operation functional unit 223a receives the counted value, operates and outputs a high-energy count rate by performing a time constant process so that the standard deviation is constant, and outputs an alert when the high-energy count rate increases beyond an acceptable set value. The low-energy count-rate-measurement functional unit 23 includes a low-window pulse-height discriminator 231, a low counter 232, and a low-energy count-rate-operation functional unit 233, the low-window pulse-height discriminator 231 receives the analog voltage pulse output from the pulse amplifier 21 and discriminates the pulse entering the window having the set low energy to output a digital pulse, and the low counter 232 counts the digital pulse at fixed cycle and outputs a counted value. In addition, the low-energy count-rate-operation functional unit 233 receives the counted value and operates and outputs a low-energy count rate by performing a moving average operation for a constant measurement time. The high counter 222 and the low counter 232 repeatedly perform set/reset for each set time, that is, fixed cycle (operation cycle) and count input pulses for a period of the fixed cycle to output a counted value. The alert-diagnosis functional unit 24 receives an alert from the high-energy count-rate-measurement functional unit 22a, receives the low-energy count rate from the low-energy count-rate-measurement functional unit 23, and determines whether or not the low-energy count rate is in a set acceptable range by performing synchronizing with alert transmission. When the low-energy count rate is in the set acceptable range, the alert-diagnosis functional unit determines that the alert is caused by fluctuation, and when the low-energy count rate increases beyond the acceptable range, the alert-diagnosis functional unit determines that the alert is caused by any of an increase in the γ ray which is a measurement target or enter of noise, and outputs results of the determination. The interface functional unit 25 receives the high-energy count rate and the alert from the high-energy count-rate-measurement functional unit 22a and results of the determination from the alert-diagnosis functional unit 24, and outputs the items in a determined order, and the display/user-operation unit 26 receives and displays each output from the interface functional unit 25 and performs operations and setting of the radiation measurement unit 2. In addition, the low-energy count rate is also input to the interface functional unit 25 from the low-energy count-rate-measurement functional unit 23. FIGS. 2A to 2C are diagrams showing windows and spectra of the radiation monitor according to Embodiment 1, and illustrate spectra observed when observation is performed by connecting a provisional multi-channel pulse height analyzer to an output of the pulse amplifier 21 in the sensitive main steam pipe monitor. Herein, the energy of the horizontal axis indicates pulse height values of a pulse wave pattern. FIG. 2A is a diagram schematically showing energy spectra in a normal state, a reference numeral a in FIG. 2A indicates background spectra, a reference numeral NL indicates a low window, and a reference numeral NH indicates a high window, respectively. FIG. 2B schematically shows energy spectra at the time of enter of noise and a reference numeral b indicates energy spectra in which noise spectra are superimposed on the background spectra a when electrostatic discharge light is generated in the radiation detector 1. In addition, FIG. 2C schematically shows energy spectra when coolant is leaked from the steam generator (SG) and radioactivity is increased and a reference c indicates spectra when a count rate of the high-energy count-rate-operation functional unit 223a is increased due to the N-16 nuclide. However, for example, in a state where the background count rate is as low as approximately 5 cpm, the alert setting point is approximately 10 cpm which slightly exceeds the alert setting level, the peak in the high-energy window NH is not clear. As shown in FIG. 2A, in a background state, a ratio between the low-energy count rate of the low-energy window NL and the high-energy count rate of the high-energy window NH is great and the low-energy count rate is several hundred times of the high-energy count rate. In addition, as shown in FIG. 2B, at the time of enter of noise, the low-energy count rate of the low-energy window NL and the high-energy count rate of the high-energy window NH are synchronously increased, and a ratio of the increased amounts (net weights) of the respective count rates is great and the low-energy count rate is several ten times of the high-energy count rate. In FIG. 2B, a reference numeral X indicates an amount of noise spectra which are superimposed on spectra in a normal state. Meanwhile, in a case where leakage from the primary coolant to the secondary coolant occurs due to leakage in the steam generator (SG), the γ ray (6.13 MeV) from the N-16 nuclide is detected, and accordingly, the high-energy count-rate-measurement functional unit 22a counts a photoelectric peak, a single-escape peak, and a double-escape peak of the γ ray which is a measurement target nuclide, as shown with a reference numeral Yin FIG. 2C, and the high-energy count rate of the high-energy window NH is increased. In addition, the low-energy count-rate-measurement functional unit 23 counts Compton scattering of the γ ray from the N-16 nuclide as shown with a reference numeral Z in FIG. 2C, the low-energy count rate of the low-energy window NL increases, but a ratio of the respective increased amount is approximately 9. A count rate m output by the high-energy count-rate-operation functional unit 223a is acquired for each fixed cycle by the following Expressions (1) to (5), when the standard deviation thereof is represented as σ, the time constant is represented as τ, the counted value is represented as M, the fixed cycle time is represented as ΔT, a value of the previous operation cycle is represented as (previous time), and a value of the current operation cycle is represented as (current time). In the descriptions hereinafter, a value of the previous operation cycle is represented as (previous time) and a value of the current operation cycle is represented as (current time).σ=1/(2mτ)1/2 (1)τ=1/(2mσ2) (2)m(current time)=m(previous time)·(1−σ)+{M(current time)/ΔT}·α (3)α=1−exp(−ΔT/τ) (4)τ=1/{2·m(previous time)·σ2} (5) That is, the count rate m output from the high-energy count-rate-operation functional unit 223a is controlled so that the standard deviation σ is constant and the time constant τ is in inverse proportion to the count rate m. It is possible to ensure desired precision by setting the standard deviation σ constant. A count rate n output from the low-energy count-rate-operation functional unit 233 is acquired for each fixed cycle by the following Expression (6), by setting the following. N: low-energy count value (fixed cycle measurement) τ (BG): time constant corresponding to the background count rate m, and it is calculated from Expression (2) based on an average value m (BG) of the count value m for a long time, for example, 24 hours in a normal state ΣN: low-energy cumulative count value (moving average cumulative time T=2τ (BG)=fixed value)n=ΣN/{2τ(BG)} (6) 2τ (BG) is set as a fixed value because a relationship of σ=1/(count rate×cumulative time)1/2=1/(count rate×2τ)1/2 is generally satisfied and the cumulative time, that is, the moving average cumulative time and 2τ are equal values. From field experiments, it has been confirmed that the time from a rise of fluctuation of a trend of the background of the high-energy count rate of the window in which the N-16 nuclide is set as a measurement target, from the average level thereof to the restoration of the fluctuation to the average level is generally 2τ, and it is found that the moving average cumulative time T=2τ is suitable as a diagnosis time. In a pressurized water reactor (PWR) plant, the sensitive main steam pipe monitor which senses leakage occurring from the primary coolant to the secondary coolant of the steam generator (SG) by monitoring a change of the N-16 nuclide by setting the N-16 nuclide as a measurement target, mainly monitors a change thereof from the background count value and can perform the measurement by matching measurement times of the high-energy count rate and the low-energy count rate in the background state, by setting the moving average cumulative time T as 2τ (BG). For example, when σ=0.1, the count rate m in the background state is 5 cpm, and n is 2,000 cpm, τ (BG) is acquired as 10 minutes from Expression (2), and therefore, 2τ (BG) becomes 20 minutes. Accordingly, the moving average cumulative time T is 20 minutes, the cumulative count value is 40000 counts, the fluctuation corresponding to the standard deviation σ is 400001/2=200 counts, and when this is divided by 20 minutes, the value is 10 cpm. Meanwhile, when the alert setting value is set as 10 cpm and a ratio between a net increase Δn of the count rate n due to leakage of the steam generator (SG) and a net increase Δm of the count rate m is set as Δn/Δm=k, a relationship of Δn=k·Δm is satisfied, when an alert is transmitted. Since k is assumed as approximately 9, for example, Δn is 45 cpm and 45 cpm/10 cpm is 4.5 σ. The high-energy count rate is increased, the alert is transmitted from the high-energy count-rate-measurement functional unit 22a, and the alert-diagnosis functional unit 24 determines whether or not the low-energy count rate is increased beyond the set acceptable range. When the low-energy count rate is in the set acceptable range and the alert is caused by the fluctuation, 4.5σ indicates the frequency in which a possibility of erroneous alert according to the calculation is once in about 11 years, under conditions of the moving average cumulative time T=20 minutes, and this means that reliability of the determination is extremely high, and an increase in fluctuation and radiation can be identified. An alert occurrence frequency due to fluctuation can be calculated, evaluated, and determined, by setting a relationship between the moving average cumulative time T and the time constant τ as lτ<T<3τ with differences of the high-energy count rate m, the low-energy count rate n, and the net increase ratio k. The alert transmission is generally caused by the statistical fluctuation according to the radiation measurement from the past experiments, and accordingly, when it is determined that the alert is caused by the fluctuation with this primary classification, the confirmation of the soundness of the apparatus to be safe, that is, on-line investigation performed by connecting a measurement device such as a provisional digital oscilloscope and multi-channel pulse height analyzer to the output of the pulse amplifier 21 and off-line investigation performed by check radiation source emission become unnecessary. As described above, in the radiation monitor according to Embodiment 1, the high-energy count-rate-measurement functional unit 22a counts the pulses entering the high-energy window NH which is set to contain a photoelectric peak, a single-escape peak, and a double-escape peak of the γ-ray (6.13 MeV) which is the N-16 nuclide, and measures the high-energy count rate by performing a time constant process so that the standard deviation is constant. The low-energy count-rate-measurement functional unit 23 counts Compton scattering of the γ ray (6.13 MeV) which is the N-16 nuclide entering the low-energy window NL, and measures the low-energy constant rate by performing a moving average operation of the high-energy count rate for a constant measurement time which is double the time constant in a background state. In addition, when the high-energy count-rate-measurement functional unit 22a outputs an alert due to an increase in the high-energy count rate, the alert-diagnosis functional unit 24 determines whether or not the low-energy count rate is increased beyond the set acceptable range and determines that the alert is caused by the fluctuation, when the low-energy count rate is in the set acceptable range. Therefore, a radiation monitor having high reliability and maintainability which can shorten the total time of a year necessary for investigation of the cause of the alert transmission with this primary classification is obtained. Next, a radiation monitor according to Embodiment 2 of the invention will be described. In Embodiment 1, the alert-diagnosis functional unit 24 identifies the statistical fluctuation of the radiation measurement which is the general cause of the alert transmission and other causes and outputs the results thereof, but in Embodiment 2, the alert-diagnosis functional unit 24 outputs the results thereof by performing secondary classification, in addition to this primary classification. FIG. 3 is a diagram showing a flow of determination of a radiation monitor according to Embodiment 2 of the invention. FIG. 3 shows a case where a noise diagnosis is added as the secondary classification of Embodiment 2 to the fluctuation diagnosis of Embodiment 1 as the primary classification. “n≦(1+p·σ)·n (BG)?” in Step S3 indicates determination of fluctuation diagnosis, n (BG) indicates an average value of the count rates n measured for a long time, p indicates a ratio of a spread of the standard deviation, and as described in Embodiment 1, when the standard deviation is set as 4.5, for example, the possibility of the erroneous determination becomes sufficiently low so as to be ignored. The configuration of the radiation monitor is the same as that in FIG. 1 and will be described with reference to FIG. 1. As shown in FIG. 3, the alert-diagnosis functional unit 24 receives the high-energy count rate m and the alert from the high-energy count-rate-measurement functional unit 22a and receives the low-energy count rate n from the low-energy count-rate-measurement functional unit 23 in Step S1. It is determined whether or not the alert is transmitted in Step S2. When the result is NO, the process returns to Step S1, and when the result is YES, a process in Step S3 as the noise diagnosis is executed and it is determined whether or not the low-energy count rate n satisfies a relationship of n≦(1+p·σ)·n (BG). When the result of the determination is YES in Step S3, it is determined that the alert is caused by the “fluctuation” in Step S4 and the result of the determination is output in Step S9. When the result of the determination is NO in Step S3, a low-energy count rate increased amount Δn and a high-energy count rate increased amount Δm are acquired and a ratio Δn/Δm thereof is further acquired in Step S5. In Step S6, it is determined whether or not a relationship of Δn/Δm≧r is satisfied as the noise diagnosis. When the result thereof is YES, it is determined that the alert is caused by the “enter of noise” in Step S7 and the result of the determination is output in Step S9. In addition, when the result of the determination is NO in Step S6, it is determined that the alert is caused by the “increase in measurement target radiation” in Step S8 and the result of the determination is output in Step S9. When the alert is output, the determination is output and the diagnosis is held, but the diagnosis is resumed by resetting the alert, for example. When an insulating material is cracked or rubbed in the radiation detector 1, charge instantaneous transfer noise is caused in a case where the insulating material is coated on a core line of a signal line, and discharge light noise is caused due to enter of the generated discharge light on a photomultiplier tube, in a case where the insulating material is close to the photomultiplier tube. In addition, when contact failure occurs in plugs of a connector or the like, contact failure noise is generated. In addition to the noises, past data is managed and confirmation is performed by experiments regarding space propagation or an effect by magnetic noise that enters from a ground wire, and as a result, Δn/Δm generally satisfies a relationship of 12<Δn/Δm, and accordingly, like the radiation monitor according to Embodiment 2, by adding the noise diagnosis based on Δn/Δm which is a ratio between the low-energy count rate increased amount Δn=n−n (BG) and the high-energy count rate increased amount Δm=m−m (BG), it is possible to identify and determine a case where the amount of the measurement target radiation is increased due to the increase of the N-16 nuclide and a case where the fluctuation of the radiation and noise are the causes, at the time of the alert transmission due to an increase of the high-energy count rate m. Therefore, by outputting the result of the determination thereof, effects of facilitating the maintaining easier and obtaining a radiation monitor having high reliability and maintainability are realized. Next, a radiation monitor according to Embodiment 3 of the invention will be described. In Embodiment 1, in the high-energy count-rate-measurement functional unit 22a, the high counter 222 counts the digital pulses output from the high-window pulse-height discriminator 221 and the high-energy count-rate-operation functional unit 223a operates and outputs the high-energy count rate by performing the time constant process so that the standard deviation is constant, based on the counted value, but in Embodiment 3, a radiation monitor for expecting high precision is obtained with a configuration of using an up-down counter, instead of the high counter. FIG. 4 is a diagram showing a configuration of the radiation monitor according to Embodiment 3 of the invention. As shown in FIG. 4, a high-energy count-rate-measurement functional unit 22b of the radiation monitor according to Embodiment 3 includes the high-window pulse-height discriminator 221, a high integration unit 224, and a high-energy count-rate-operation functional unit 223b, and the high integration unit 224 includes an up-down counter 2241, a negative feedback pulse generation circuit 2242, and an integration control circuit 2243. The high-window pulse-height discriminator 221 receives the analog voltage pulse output from the pulse amplifier 21 and discriminates the pulse entering the window having the set high energy to output a digital pulse, the up-down counter 2241 receives the digital pulse output from the high-window pulse-height discriminator 221 through an up input, the negative feedback pulse generation circuit 2242 generates a feedback pulse at a repetition frequency so as to respond the output of the up-down counter 2241 with a primary delay of the time constant and inputs the feedback pulse to a down input of the up-down counter 2241. Herein, the up-down counter 2241 includes the up input and the down input, in which the up input proceeds the counting and the down input restores the counting process and outputs an addition and subtraction integration value as a result of addition and subtraction. A signal pulse of a detector line which is the same as that of the high counter of Embodiment 1 is input to the up input, the negative feedback pulse is input to the down input, and addition and subtraction integration is performed consecutively without resetting. Therefore, the repetition frequency of the feedback pulse responding at the time constant with a primary delay is in equilibrium with respect to the repetition frequency of the input pulse, the inputs are switched to each other with an addition and subtraction integrated value in this state, and a stabilized state is obtained with oscillation for only a weighed amount of 1 pulse. The integration control circuit 2243 performs weighing when the up-down counter 2241 performs the counting in accordance to the standard deviation of the count rate, and the high-energy count-rate-operation functional unit 223b operates the count rate m by the following Expressions (7) to (9) so that the standard deviation σ is constant based on the addition and subtraction integrated value Q of the up-down counter 2241. In addition, the negative feedback pulse generation circuit 2242 generates the feedback pulse based on the addition and subtraction integrated value Q. The current operation cycle is represented as (current time).γ=2σ2=1/{m(current time)·τ(current time)}=2−λ\|n2 (7)β=11−λ (8)m(current time)=exp{γ·Q(current time)} (9) herein, γ, λ, and β are constants. When a relationship of β=0 is set as a reference by Expression (8), the addition and subtraction integrated value Q (current time) when λ is 11, responds by an increase or a decrease of 1 count with respect to the input of 1 count, and the addition and subtraction integrated value Q (current time) when β is 2 and λ is 9 responds an increase or a decrease of 4 counts with respect to the input of 1 count. In addition, the addition and subtraction integrated value Q (current time) when β is 4 and λ is 7 responds an increase or a decrease of 16 counts with respect to the input of 1 count, and the addition and subtraction integrated value Q (current time) when β is 6 and λ is 5 responds an increase or a decrease of 64 counts with respect to the input of 1 count. That is, when the count rate m (current time) is set constant, the response time t (current time) depends on the weighing of the counts with respect to the input of the up-down counter 2241. The other configurations and operations are the same as those in Embodiment 1 and therefore, the overlapped description will be omitted by setting the same reference numerals. As described above, the high counter 222 of the radiation monitor according to Embodiment 1 generates loss time according to the resetting, however, the up-down counter 2241 of the radiation monitor according to Embodiment 3 does not need the resetting and consecutively performs adding and subtraction integration, and therefore, it is possible to expect excellent linearity, that is, high precision, to the point of the high count rate. Next, a radiation monitor according to Embodiment 4 of the invention will be described with reference to FIGS. 5 and 6. In Embodiment 1, the radiation measurement unit 2 performs the fluctuation diagnosis based on the low count rate, and in Embodiment 2, in the same manner, the radiation measurement unit 2 performs the fluctuation diagnosis and the noise enter diagnosis based on the low count rate. As shown in FIG. 5, in Embodiment 4, the radiation measurement unit is configured with a high-energy radiation measurement unit 3 which is first radiation measurement means and a low-energy radiation measurement unit 4 which is second radiation measurement means, the analog voltage pulses output from the radiation detector 1 are respectively input to the high-energy radiation measurement unit 3 and the low-energy radiation measurement unit 4, and the high-energy radiation measurement unit 3 operates in the same manner as the high-energy count-rate-measurement functional unit 22a of Embodiment 1 or the high-energy count-rate-measurement functional unit 22b of Embodiment 3, to output the high-energy count rate and the alert. In addition, the low-energy radiation measurement unit 4 operates in the same manner as the high-energy radiation measurement unit 3 and outputs the low-energy count rate, with a configuration in which the high-window pulse-height discriminator 221 of the high-energy count-rate-measurement functional unit 22a of Embodiment 1 or the high-energy count-rate-measurement functional unit 22b of Embodiment 3 is switched with the low-window pulse-height discriminator 231. The low-energy radiation measurement unit 4 has a function of the transmission of the alert, if necessary. The measurement energy range of the low-energy radiation measurement unit 4 is set so as to contain peak spectra of radioactive rare gas which is an emission management target and main Compton scattering spectra, as shown in FIG. 6. Accordingly, in the standard deviation σ in Expression (1), when the ratio of the standard deviation of the low-energy radiation measurement unit 4 to the high-energy radiation measurement unit 3 is set as ¼, for example, in a case where the standard deviation of the high-energy constant rate is 0.1, the standard deviation of the low-energy constant rate becomes 0.025. When the background count rate of the high-energy count rate is 5 cpm and the background count rate of the low-energy count rate is 2,000 cpm, the time constant of the high-energy count rate is 10 minutes and the time constant of the low-energy count rate is 0.4 minutes from Expression (2), and it is possible to expect emission of radioactive rare gas in a preferred state with a balance between the fluctuation and the response. A diagnosis apparatus 5 shown in FIG. 5 includes an alert diagnosis unit 51 and a display unit 52, and receives high-energy count rate and the alert from the high-energy radiation measurement unit 3 and the low-energy count rate from the low-energy radiation measurement unit 4. The alert diagnosis unit 51 includes a fluctuation diagnosis functional unit 511 and a noise diagnosis functional unit 512, operates in the same manner as the alert-diagnosis functional unit 24 of Embodiment 1 or Embodiment 2, outputs a result of the fluctuation diagnosis from the fluctuation diagnosis functional unit 511, and outputs a result of the noise enter diagnosis from the noise diagnosis functional unit 512. The display unit 52 simultaneously displays the result of diagnosis of the alert diagnosis unit 51 and the trend of the high-energy count rate and the low-energy count rate. When the time constant at the background level of the high-energy radiation measurement unit 3 is set as τ1 and the time constant at the background level of the low-energy radiation measurement unit 4 is set as τ2, a relationship of τ1>>τ2, and by setting the moving average time τ2 of the diagnosis apparatus 5 to satisfy a relationship of 1×τ1<τ2<3τ×1 and preferably a relationship of T=2×τ1, it is possible to properly identify a case where the alert transmission is caused by the fluctuation, in the same manner as the noise diagnosis of Embodiment 1. In the display of the trend, a horizontal axis indicates the time, a left part of a vertical axis shows the high-energy count rate and a right part thereof shows the moving average value of the low-energy count rate in the screen, linear and logarithm can be desirably selected as scales thereof and the displaying is set so as to be performed by expanding or contracting the range, and accordingly, it is possible to determine the cause of the indication increase visually. FIG. 6 shows a relationship between the low-energy window and the energy of radiation of the rare gas nuclide and a positional relationship with the high-energy window, and accordingly, in FIG. 6, Xe-135, Ar-41, Kr-85, Kr-87, and Kr-88 indicate rare gas nuclides, and Y1, Y2, and Y3 respectively shows a double-escape peak, a single-escape peak, and a photoelectric peak of the N-16 nuclide. In the display of the low-energy radiation measurement unit 4 and the display of the diagnosis apparatus 5, the low-energy count rate has a range with a low concentration of released radioactivity by the radiation monitor of the invention and has a range with a high concentration thereof by another radiation monitor, and may be displayed as a dose equivalent rate, for example, for matching the units of the measured values of the low range and the high range. In addition, the diagnosis apparatus 5 may be integrated with a calculator system of the plant. Hereinabove, Embodiment 1 to Embodiment 4 of the invention have been described, but the invention is not limited to those embodiments, and each embodiment can be freely combined with each other or modifications and omissions of the embodiments can be suitably performed within a scope of the invention.
	description	This invention relates generally to radiation apparatuses and methods. In particular, various embodiments of multileaf collimators with alternating trapezoidal leaf geometry design are described. Multileaf collimators (MLCs) are widely used in radiotherapy machines to support various forms of treatment including 3D conformal radiation therapy (3D-CRT), intensity-modulated radiotherapy (IMRT), volumetric modulated arc therapy (VMAT), etc. An MLC includes a plurality pairs of beam-blocking leaves arranged in opposing banks. Individual beam-blocking leaves can be independently moved in and out of a radiation beam to block or modify the beam. In use, selected beam-blocking leaves can be positioned in the radiation beam, forming one or more apertures through which the unblocked radiation beam passes. The aperture(s) define(s) the shape of the radiation beam directed to a treatment field at an isocenter. Tolerance or gap between adjacent beam-blocking leaves in an MLC exists or is provided to allow dynamic linear or longitudinal movement of the leaves. The interleaf gap or tolerance can be a source of radiation leakage in an MLC. The interleaf tolerance may also cause the leaves to flop when the MLC or the gantry supporting the MLC rotates such that the amount of interleaf leakage is unpredictable. Controlling the amount of interleaf leakage for all gantry and collimator angles is also important so that the radiation delivered by the radiation system to the target is of acceptable quality. To mitigate MLC interleaf leakage, various leaf designs are developed, including “tongue in groove” designs in which steps or similar geometries are provided on the leaf sides so that leaf materials mutually overlap between leaves. Manufacturing of beam-blocking leaves with a “tongue in groove” design can be very expensive. Further, while a “tongue in groove” design may reduce interleaf leakage, it may lead to undesirable underdose effects when MLC treatment fields are combined. U.S. Pat. No. 7,742,575 B2 discloses an MLC in which the beam-blocking leaves are held to be shifted or orientated such that the sides or faces of the leaves align with a convergence point which offsets the radiation source. As such, beams from the radiation source would strike the beam-blocking leaves at an angle, avoiding the gap between the adjacent leaves through which radiation could pass uninterrupted. The asymmetrical shift of the beam-blocking leaves from the radiation source creates variations in penumbra and resolution across the entire field. Embodiments of this disclosure provide for a multileaf collimator (MLC) having an alternating trapezoidal leaf geometry design. The novel leaf geometry design can reduce MLC interleaf leakage, provide predictability of interleaf leakage regardless of dynamic rotation of the MLC, and maintains uniform side leaf penumbra across the treatment field. The flat sided, trapezoidal leaf geometry design allows for reduction of the manufacturing costs of each individual leaf, and thus the MLC. In one embodiment, a multileaf collimator comprises a plurality of beam-blocking leaves of a first type and a plurality of beam-blocking leaves of a second type. Each of the beam-blocking leaves of the first type has a trapezoidal geometry viewed in the longitudinal moving direction comprising a first lateral side, a second lateral side, a wider end and a narrower end with the wider end being proximal to a source. Each of the beam-blocking leaves of the second type has a trapezoidal geometry viewed in the longitudinal moving direction comprising a first lateral side, a second lateral side, a wider end and a narrower end with the wider end being distal to the source. The beam-blocking leaves of the first type are alternatingly arranged with the beam-blocking leaves of the second type side by side. The first lateral sides of the trapezoidal geometry of the plurality of beam-blocking leaves of the first type may align to converge to a first point offset from the source. The second lateral sides of the trapezoidal geometry of the plurality of beam-blocking leaves of the first type may align to converge to a second point offset from the source opposite to the first point. The first lateral sides of the trapezoidal geometry of the plurality of beam-blocking leaves of the second type may align to converge to the second point, and the second lateral sides of the trapezoidal geometry of the plurality of beam-blocking leaves of the second type may align to converge to the first point. The first and second lateral sides of the plurality of beam-blocking leaves of the first type and the first and second lateral sides of the plurality of beam-blocking leaves of the second type may be substantially flat. In one embodiment, a multi-level multileaf collimator (MLC) comprises a first MLC in a first level distal to a source and a second MLC in a second level proximal to the source. The second MLC comprises a plurality of beam-blocking leaves of a first type and a plurality of beam-blocking leaves of a second type. Each of the beam-blocking leaves of the first type has a trapezoidal geometry viewed in the longitudinal moving direction comprising a first lateral side, a second lateral side, a wider end and a narrower end with the wider end being proximal to a source. Each of the beam-blocking leaves of the second type has a trapezoidal geometry viewed in the longitudinal moving direction comprising a first lateral side, a second lateral side, a wider end and a narrower end with the wider end being distal to the source. The beam-blocking leaves of the first type are alternatingly arranged with the beam-blocking leaves of the second type side by side. The plurality of beam-blocking leaves of the first MLC may be longitudinally movable in a direction substantially parallel with the longitudinal moving direction of the beam-blocking leaves of the second MLC. Each of the beam-blocking leaves of the first and second types of the second MLC may laterally offset a beam-blocking leaf of the first MLC. In one embodiment, an apparatus includes a source of radiation, and a multileaf collimator comprising a plurality of beam-blocking leaves of a first type and a plurality of beam-blocking leaves of a second type. Each of the beam-blocking leaves of the first type has a trapezoidal geometry viewed in the longitudinal moving direction comprising a first lateral side, a second lateral side, a wider end and a narrower end with the wider end being proximal to a source. Each of the beam-blocking leaves of the second type has a trapezoidal geometry viewed in the longitudinal moving direction comprising a first lateral side, a second lateral side, a wider end and a narrower end with the wider end being distal to the source. The beam-blocking leaves of the first type are alternatingly arranged with the beam-blocking leaves of the second type side by side. The source of radiation may be a source of x-rays, a source of gamma rays, a source of protons, or a source of heavy ions. This Summary is provided to introduce selected aspects and embodiments of this disclosure in a simplified form and is not intended to identify key features or essential characteristics of the claimed subject matter, nor is it intended to be used as an aid in determining the scope of the claimed subject matter. The selected aspects and embodiments are presented merely to provide the reader with a brief summary of certain forms the invention might take and are not intended to limit the scope of the invention. Other aspects and embodiments of the disclosure are described in the section of Detailed Description. Referring to FIGS. 1-5, various embodiments of multileaf collimators (MLCs) with an alternating trapezoidal leaf geometry design will now be described. FIG. 1 is a simplified illustration of a radiation system 100 including an MLC 110 according to embodiments of the disclosure. As shown, the radiation system 100 may include a radiation source 102 configured to produce a beam 103 of radiation such as photons, electrons, protons, or other types of radiation. For example, the radiation source 102 may include a metallic target configured to produce a beam of x-rays upon impingement of electrons. The radiation system 100 may include various beam shaping components such as a primary collimator 104 and optionally a secondary collimator 106 to generally limit the extent of the beam 103 as it travels away from the radiation source 102 toward an isocenter plane 108. An MLC 110 such as a multi-level MLC is disposed between the radiation source 102 and the isocenter plane 108 to further shape the beam 103, as indicated by the shaped field 112 in the isocenter plane 108, according to a general use embodiment of the MLC 110. The MLC 110 may rotate about the beamline or axis 109 passing through the radiation source 102, placing the MLC 110 in various orientations. The radiation source 102, primary collimator 104, secondary collimator 106, and the MLC 110 may be enclosed in a treatment head (not shown), which can be rotated by a gantry (not shown) about an axis such as a horizontal axis 111. Thus, the radiation system 100 can deliver treatment beams to a target in the isocenter plane 108 from various angles. The shape, size, and/or intensity of the beam 103 can be adjusted or dynamically adjusted by the MLC 110 as the beam angle is stepped or swept around the target. The MLC 110 may be a single level MLC or a multi-level MLC as shown. By way of example, the MLC 110 may include a first MLC 120 in a first level distal to the radiation source 102 and a second MLC 130 in a second level proximal to the radiation source 102. As used herein, the term “multileaf collimator” or “MLC” refers to a collection of a plurality of beam-blocking leaves each of which can be longitudinally moved in and out of a beam 103 to modify one or more parameters of the beam 103 such as the beam shape, size, energy, or intensity etc. Each beam-blocking leaf may be driven by a motor with a lead screw or other suitable means. The beam-blocking leaves may be arranged in pairs. The beam-blocking leaves of each pair may be brought in contact or retracted from each other to close or open a path for a radiation beam to pass through the MLC 110. The beam-blocking leaves may be arranged in opposing banks and supported by a frame, box, carriage or other support structure, which has features allowing the individual beam-blocking leaves to extend into and retract from the beam 103. The frame, box, carriage or other support structure can be further moved or translated in addition to the individual leaf travel. As shown in FIG. 1, the first MLC 120 and the second MLC 130 may be arranged such that the moving direction of individual beam-blocking leaves of the first MLC 120 and the second MLC 130 are generally in parallel. For example, as shown in FIG. 1 the beam-blocking leaves 122 of the first MLC 120 in the first level are longitudinally movable in the x-direction, and the beam-blocking leaves 132 of the second MLC 130 in the second level are also longitudinally movable in the x-direction. Alternatively, the first MLC 120 and the second MLC 130 may be arranged such that the moving direction of the beam-blocking leaves 122 of the first MLC 120 is non-parallel e.g. perpendicular to the moving direction of the beam-blocking leaves 132 of the second MLC 130. The first MLC 120 and the second MLC 130 may be arranged such that the beam-blocking leaves 132 of the second MLC 130 may laterally offset the beam-blocking leaves 122 of the first MLC 120 in a beam's eye view, or as viewed in a direction from the radiation source 102. FIG. 2 is a cross-sectional view of a portion of the multi-level MLC 110 of FIG. 1 taken along line A-A, showing the lateral offset arrangement of the beam-blocking leaves of the multi-level MLC 110. As shown, a beam-blocking leaf 132 of the second MLC 130 in the second level offsets a beam-blocking leaf 122 of the first MLC 120 in the first level as viewed from the radiation source 102. By way of example, a beam-blocking leaf 132 of the second MLC 130 may offset a beam-blocking leaf 122 of the first MLC 120 by substantially half a beam-blocking leaf. Alternatively, a gap between two adjacent beam-blocking leaves 132 of the second MLC 130 in the second level may be positioned substantially at the middle of a beam-blocking leaf 122 of the first MLC 120. The lateral offset arrangement of beam-blocking leaves in different levels provides for leaf projections that are also offset at the isocenter plane 108. Therefore, the lateral offset arrangement of beam-blocking leaves may provide for substantially an equivalent of doubling MLC definition, or improving the resolution to half as compared to the definition of a single level MLC 110 with beam-blocking leaves of the same physical width. In some embodiments, three or more MLCs may be arranged in three or more levels such that each beam-blocking leaf at a level may offset e.g. by ⅓ or 1/n of a leaf width as projected at the isocenter plane 108 where n is the number of the MLCs. U.S. Pat. No. 8,637,841 issued on Jan. 28, 2014 to the common assignee entitled “Multi Level Multileaf Collimators” describes various embodiments of multi-level MLCs, the disclosure of which is incorporated herein by reference in its entirety. In some embodiments, the MLC 110 may include beam-blocking leaves having a trapezoidal geometry viewed in the beam-blocking leaf longitudinal moving direction. As used herein, the term “trapezoidal geometry” or its grammatic equivalent refers to a geometry including a wider end and a narrower end parallel to each other and two lateral sides connecting the wider end 142 and the narrower end 144. In some embodiments, the lateral sides of the beam-blocking leaves are substantially flat. FIG. 3 depicts a bank of beam-blocking leaves 140 of an exemplary MLC 110 having a trapezoidal geometry design according to embodiments of the disclosure. As shown, a beam-blocking leaf 140 of a trapezoidal geometry design may include a wider end 142 and a narrower end 144 parallel to each other, and a first or left lateral side 146 and a second or right lateral side 148 connecting the wider end 142 and the narrower end 144. In some specific embodiments, the plurality of beam-blocking leaves 140 of trapezoidal geometry design may be arranged such that the beam-blocking leaves 140 whose wider ends 142 are proximal to the radiation source 102 alternate with the beam-blocking leaves 140 whose wider ends 142 are distal to the radiation source 102. For ease of description, in the Detailed Description and Claims, the term “beam-blocking leaf of a first type” or its grammatic equivalent may be used to refer to a beam-blocking leaf 140 having a trapezoidal geometry with the wider end 142 being arranged proximal to the radiation source 102. The term “beam-blocking leaf of a second type” or its grammatic equivalent may be used to refer to a beam-blocking leaf 140 having a trapezoidal geometry with the wider end 142 being arranged distal to the radiation source 102. Therefore, according to embodiments of the disclosure, the MLC 110 may include a plurality of beam-blocking leaves 140 of a first type and a plurality of beam-blocking leaves 140 of a second type, where the beam-blocking leaves 140 of the first type are alternatingly arranged with the beam-blocking leaves 140 of the second type side by side. FIG. 4 illustrates a sub-set of beam-blocking leaves 140 in an exemplary MLC 110 according to embodiments of the disclosure, showing the alternating arrangement of the beam-blocking leaves 140. As shown, each of the beam-blocking leaves 140 has a trapezoidal geometry design as viewed in the leaf longitudinal moving direction. The MLC 110 includes beam-blocking leaves of the first type 140a, with the wider end 142 being proximal to the radiation source 102, and beam-blocking leaves of the second type 140b, with the wider end 142 being distal to the radiation source 102. The beam-blocking leaves of the first type 140a alternate with the beam-blocking leaves of the second type 140b side by side. According to embodiments of the disclosure, the beam-blocking leaves of the first type 140a and the beam-blocking leaves of the second type 140b can be designed and constructed such that when fitted, the first or left lateral sides 146 of the beam-blocking leaves of the first type 140a align to converge to a first point offset from the radiation source 102, and the second or right lateral sides 148 of the beam-blocking leaves of the first type 140a align to converge to a second point offset from the radiation source 102. For the beam-blocking leaves of the second type 140b, the first or left lateral sides 146 of the beam-blocking leaves 140b may align to converge to the second point offset from the radiation source 102 and the second or right lateral sides 148 align to converge to the first point offset from the radiation source 102. Referring to FIG. 5, for example, the left or first lateral sides 146a of the plurality of beam-blocking leaves of the first type 140a align to converge to a point 102a, which offsets the radiation source 102 at a distance (−X). The right or second lateral sides 148a of the plurality of beam-blocking leaves of the first type 140a align to converge to a point 102b which offsets the radiation source 102 at a distance (+X). Still referring to FIG. 5, the left or first lateral sides 146b of the plurality of beam-blocking leaves of the second type 140b, which face or are adjacent to the right or second lateral sides 148a of the beam-blocking leaves of the first type 140a, align to converge to the second point 102b, which offsets the radiation source 102 at a distance (+X). The right or second lateral sides 148b of the plurality of beam-blocking leaves of the second type 140b, which face or are adjacent to the left or first sides 146a of the beam-blocking leaves of the first type 140a, align to converge to the first point 102a, which offsets the radiation source 102 at a distance (−X). The first converging point 102a may offset the radiation source 102 at a distance substantially equal to the distance that the second converging point 102b offsets the radiation source 102 at an opposite side. The alternating trapezoidal geometry design of the beam-blocking leaves allows the flat side surfaces of adjacent leaves to be slightly “off-focus” relative to the radiation source 102, hence change the angle of the path or gap between adjacent beam-blocking leaves that a radiating beam from the radiation source 102 would pass. As shown in FIG. 5, the left lateral side 146a of beam-blocking leaf 140a forms an angle with a line passing the middle point (M) of the lateral side 146a and the radiation source 102. Similarly, the right lateral side 148a of beam-blocking leaf 140a forms an angle with a line passing the middle point (M) of the right lateral side 148a and the radiation source 102. These angles allow a radiation beam from the radiation source 102 to be attenuated by both a top corner of a beam-blocking leaf and a bottom corner of an adjacent beam-blocking leaf, thereby reducing interleaf leakage. The plurality of beam-blocking leaves 140 can be designed and constructed such that these off-focus angles are consistent for each of the beam-blocking leaves regardless of the location of the leaf in the leaf bank, providing symmetrical penumbra and resolution for the entire treatment field. The alternating trapezoidal leaf geometry design provides for improved packing of drive motors for the beam-blocking leaves because it can slit half of the motors to the proximal part of the leaf bank and half of the motors to the distal part of the leaf bank. This provides for more room for the leaf drive system including motors, lead screws etc., allowing for a more robust and reliable design. The alternating trapezoidal leaf geometry design can also reduce the costs of constructing an MLC because the boxes or support structures for both banks of the beam-blocking leaves can be made identical, reducing the number of parts and ultimately reducing the costs. Various embodiments of multileaf collimators have been described with reference to the figures. It should be noted that some figures are not necessarily drawn to scale. The figures are only intended to facilitate the description of specific embodiments, and are not intended as an exhaustive description or as a limitation on the scope of the disclosure. Further, in the figures and description, specific details may be set forth in order to provide a thorough understanding of the disclosure. It will be apparent to one of ordinary skill in the art that some of these specific details may not be employed to practice embodiments of the disclosure. In other instances, well known components or process steps may not be shown or described in detail in order to avoid unnecessarily obscuring embodiments of the disclosure. All technical and scientific terms used herein have the meaning as commonly understood by one of ordinary skill in the art unless specifically defined otherwise. As used in the description and appended claims, the singular forms of “a,” “an,” and “the” include plural references unless the context clearly dictates otherwise. The term “or” refers to a nonexclusive “or” unless the context clearly dictates otherwise. Further, the term “first” or “second” etc. may be used to distinguish one element from another in describing various similar elements. It should be noted the terms “first” and “second” as used herein include references to two or more than two. Further, the use of the term “first” or “second” should not be construed as in any particular order unless the context clearly dictates otherwise. Various relative terms such as “upper,” “above,” “top,” “over,” “on,” “below,” “under,” “bottom,” “higher,” “lower,” “left,” “right” or similar terms may be used herein for convenience in describing relative positions, directions, or spatial relationships in conjunction with the drawings. The use of the relative terms should not be construed as to imply a necessary positioning, orientation, or direction of the structures or portions thereof in manufacturing or use, and to limit the scope of the invention. Those skilled in the art will appreciate that various other modifications may be made. All these or other variations and modifications are contemplated by the inventors and within the scope of the invention.
	summary
	abstract	Garment (10) for sporting activity, in particular a suit for motorcyclists, which is made of breathable fabric in selected regions and comprises a protective shield (40). At least the bottom portion (44) of the shield is removably inserted underneath a layer (58) of puckered elastic material fixed to the breathable fabric.
	claims	1. An method for forming a liquid crystal alignment layer in a thin-film on a substrate, comprising:a step for preparing a substrate whereon a thin-film is formed that will be a liquid-crystal alignment layer;a step for placing mask, having a reflective face on the substrate side, between the substrate and an ion source; anda step for conveying the substrate, irradiating the thin-film on the substrate with an ion beam emitted from the ion source, and reflecting the ion beam between the thin-film and the reflective face of the mask, reflecting the ion beam in the orientation direction from the reflective face of the mask;wherein:the traveling direction of the substrate, in which the substrate is conveyed, is the same direction as the direction in which the ion beam is illuminated. 2. The alignment layer-forming method according to claim 1, wherein the incident angle of the ion beam onto the thin-film is varied by reflecting the ion beam with the reflective face of the mask such that the alignment layer creates a predetermined pre-tilt angle. 3. The alignment layer-forming method according to claim 1, wherein the reflective face of the mask is formed such that the spread angle of the ion beam is corrected.
055481256	summary	The present invention relates to radiation protective gloves and in particular radiation protective gloves for surgical or medical use and processes for their manufacture. Surgeons and other medical personnel are often involved in medical procedures such as diagnostic, detection or guidance procedures in which their hands are exposed to radiation such as X-rays. In many of these procedures the field of operation is irradiated with X-rays so that the surgeon or other personnel can carry out the procedure using a fluoroscopic viewing screen. In diagnostic procedures using X-rays a radiologist may have to hold a patient such as an infant or in the case of veterinary work an animal to restrain the movement thereof. The dose of radiation received by a patient in any of these procedures will normally be well below the non-acceptable levels. Surgical or medical personnel who frequently carry out these procedures, however, may be exposed to radiation above the acceptable dose level. It is therefore desirable that these personnel wear protective gloves during the above procedures to limit or attentuate the amount of radiation received by the hands. Radiation protective gloves containing lead or lead oxide fillers are known in the art. Lead compounds, however, are toxic materials. Furthermore, gloves containing lead compounds can mark surfaces, for example, with a black mark. In addition it has been found that gloves with sufficient wall thickness or filler content to provide good radiation protection tend to be inflexible thereby making the gloves tiresome to wear and difficult to use by the wearer when handling instruments. U.S. Pat. No. 5,001,354 discloses radiation protective gloves prepared by a latex dipping in a polymeric mixture comprising a dispersion of natural rubber latex and up to 20% by volume of tungsten filler which are capable of absorbing 50 to 80% incident radiation generated at voltages of 60 to 100 kVp. With gloves having the highest tungsten loading describe, the radiation absorbing capacity at higher radiation rates is limited. The difficulty associated with the use of natural rubber latex-filler dispersion disclosed in the process of U.S. Pat. No. 5,001,354 is that even at the relatively lower filler contents disclosed therein has to be continuously agitated by a complex arrangement of pumps to maintain the tungsten filler in suspension, As a consequence it has been found almost impossible to prepare latex rubber dispersions with tungsten filler content higher than 20% by volume to provide gloves with a higher radiation protection than that given in the hereinabove US Patent because of the extremely fast settling rate of the high specific gravity tungsten filler. It would be desirable to have gloves with an even higher radiation absorbing capacity to limit the effects of radiation exposure on the wearers' hands, It has now been found possible to achieve protection at higher radiation levels than with flexible gloves containing a higher tungsten filler content than 20% by volume. Such gloves can be made by relatively simple processes using a flexible polymer. Accordingly the present invention provides a radiation protective glove for surgical or medical use comprising a layer of flexible polymer containing at least 25% by volume of particulate tungsten material and having a radiation absorbing capacity equivalent to that of at least 0.13 mm thickness of lead. Gloves of the invention are preferably made of a flexible synthetic polymer. In another aspect, therefore, the present invention provides a radiation protective glove for surgical or medical use comprising a layer of flexible synthetic polymer containing at least 25% by volume of particulate tungsten material and having a radiation absorbing capacity equivalent to that of 0.13 mm thickness of lead. The gloves of the invention will normally be used in situations where the wearer is exposed to X-rays generated at voltages up to 150 KVP. The filled polymer layer of the gloves will have a radiation absorbing capacity equivalent of at least 0.13 mm thickness of lead, more suitably at least 0.25 mm thickness of lead and preferably a radiation absorbing capacity of at least 0.35 mm thickness of lead. The lead thickness equivalent of a tungsten filled layer of a glove of the invention can be obtained by measuring the % transmission through a sample layer of an x-ray beam generate at 60 KVP and comparing it with the % transmission of a similar x-ray beam through a different thickness of lead foil. % absorption or attenuation the radiation for a layer can be then obtained by subtracting the % transmission value from 100%. FIG. 3 of the drawings shows a graph of % transmission versus lead thickness for x-ray beams generated at voltages of 60, 80, 100 and 120 KVP. FIG. 3 indicates that a layer with a lead thickness equivalent of at least 0.13 mm has a % absorption of about 90% for x-rays generated at 60 KVP and in excess of 80% for x-rays generated at 100 KVP. Furthermore, it has been found that the tungsten filled polymer layers of gloves of the invention enhibit higher lead thickness equivalents with x-ray beams generated at higher voltages than that of 60 KVP. Gloves of the invention are therefore capable of absorbing well in excess of 80% of the incident radiation at 60 to 100 KVP. Gloves of the invention are capable of absorbing more suitably at least 85%, desirably at least 90% and preferably at least 95% of the incident radiation at 60 to 100 KVP. Gloves of the invention therefore can provide greater protection to x-rays than the gloves disclosed in the hereinbefore mentioned prior art United States patent. The amount of particulate tungsten material in the polymer layer of the gloves of the invention can be adapted to obtain a flexible layer with the desired radiation absorption capacity. Such an amount will be at least 25% by volume and can favourably be at least 30% by volume and can preferably be at least 40% by volume. Similarly the amount of particulate tungsten material in the polymer layer of the gloves of the invention can suitably be less than 90% by volume, more suitably less than 70% by volume and can preferably be less than 50% by volume. Apt polymer layers for use in the invention contain 30% to 60% by volume and preferably 35% to 55% by volume of particulate tungsten material. The tungsten material containing polymer layer of the gloves of the invention will not contain any holes which would allow the direct passage of x-rays. Surgical gloves of the invention will also be impermeable to aqueous liquids and bacteria to provide a barrier therefor. The thickness of this layer can suitably be less than 1.5 mm, favourably be less than 1.00 mm and can preferably be less than 0.8 mm. Similarly the thickness of the polymer layer can be suitably greater than 0.1 mm and can preferably be greater than 0.2 mm. Apt polymer layers for use in the invention have a thickness of suitably 0.1-1.3 mm and preferably 0.2-1.0 mm. The thickness of the tungsten material containing polymer layer can be adapted to provide a chosen radiation protection level (expressed as equivalent to a lead thickness) at a given filler volume percentage. It is believed that gloves of the invention can advantageously provide a level of radiation protection equivalent to 0.5 mm of lead using glove polymer layer of less than 1 mm thick at particulate tungsten material loading of 40% by volume. Apt gloves of the invention having a radiation protection level equivalent to 0.25 to 0.35 mm of lead can be provided using a 0.5-0.7 mm thick polymer layer containing 40% by volume of particulate tungsten material. Tungsten material suitable for use in the gloves of the invention include tungsten metal and chemically inert compounds thereof such as tungsten oxide and tungsten carbide. However, a higher volume percentage of tungsten compound in the gloves is required to get the same radiation absorption protection as that of tungsten metal. Tungsten materials advantageously have a higher specific gravity and a higher radiation absorption per unit thickness than that of lead material. As a consequence a layer containing tungsten material can provide higher relative radiation absorption and therefore higher radiation protection than that of similar layer containing the same volume percentage of lead material. Furthermore as herein before mentioned the tungsten filled polymer layer can provide higher than expected relative radiation absorption and protection from x-rays generated at voltages in excess of 60 KVP. The tungsten material used in the invention will be in a particulate form such as a powder. The tungsten material can have a particulate size of suitably less than 20 .mu.m, favourably less than 10 .mu.m and preferably less than 1 .mu.m for example 0.5 to 0.9 .mu.m. The tungsten containing flexible polymer layer of the glove of the invention should be sufficiently flexible to enable the wearer to bend the finger portions of the glove without undue force, to hold instruments therewith and preferably also to obtain a sense of "touch" or "feel" through the walls of the glove. Suitable flexible polymers for use in the invention can include any of the pharmaceutically acceptable and water insoluble synthetic polymers capable of forming flexible layers for use in gloves. Such polymers include elastomeric polymers ie. elastomers and plasticised non-elastomeric polymers. Favoured flexible polymers however are elastomeric polymers. Suitable elastomers include those comprising natural rubber, butadiene homopolymers and its copolymers with styrene, isobutylene-isoprene copolymers, ethylene-propylene and ethylene-propylene-diene copolymers, polybutadiene acrylate, synthetic polyisoprene, polydimethylsiloxane and thermoplastic elastomers such as polyester-urethane, polyether-urethane, polyether-amide polyether-ester and A-B-A type block copolymers where A is styrene and B is butadiene, isoprene or ethylene butylene. Aptly the flexibility of the polymer material employed in the gloves of the invention is at least 0.25 mm, more aptly at least 0.35 mm, and preferably at least 0.45 mm when determined by the following bend test method. A list rig comprised two bars, 5 mm long, spaced 5 mm apart. The bars had an inverted V-shape to provide loading surfaces. A 13 mm.times.3 mm strip of polymer material was draped over the bars and a load of 100 mmN applied to the centre of the material for 2 minutes. The deflection of the material under the load was measured employing a Perkin-Elmer Thermo-Mechanical Analyzer. The deflection in millimeters is expressed as the flexibility. Favoured elastomers include natural rubber, ethylene-propylene copolymers rubbers (EPM) and ethylene-propylene copolymers rubbers (EPDM) containing diene side chains derived from monomers such as 1, 4, hexadiene, dicyclopentadiene or ethylidenenorbornene monomers. Flexible polymers such as elastomers for use in the invention can advantageously be cross-linked or cured to render the glove layer or layers tougher. The presence of the pendant sites of unsaturation in EPDM rubbers enables these rubbers to be cross-linked or cured by conventional sulphur based rubber vulcanising systems. A layer of flexible polymer such as an elastomer used in the invention can optionally contain a plasticiser to render the glove layer or layers more flexible. The EPM and EPDM rubbers can be readily plasticised by hydrocarbon oils such as aliphatic hydrocarbon oils to advantageously provide layers with very good flexibility. The EPM and EPDM rubber layer used in the invention can suitably contain up to 50% by weight of hydrocarbon oil. Suitable plasticised non-elastomeric polymers include plasticised vinyl chloride polymers and copolymers. A flexible polymer layer used in the invention can optionally contain optionally up to 25% by weight of a filler for example to reinforce the glove layer. The tungsten containing polymer layer of a glove of the invention can be provided with a protective coating of a flexible polymer such as an elastomer on its inner or outer surface and on both such surfaces. Such a coating can suitably be less than 100 .mu.m thick and can preferably be less than 75 .mu.m thick. Apt coatings are 10 to 50 .mu.m thick. Such protective coatings are preferably on the finger or palm portions of the glove. The tungsten containing polymer layer may comprise a flexible reinforcing layer to improve the tear and puncture resistance of the polymer layer. Suitable reinforcing layers include films of a polymer such as polyurethane, polyethylene, ethylene-vinyl-acetate copolymer, non-woven fabrics or plastics nets. The reinforcing layer may be laminated to the surface of the filled polymer layer or included within the layer. In a further aspect the present invention provides a process of forming a radiation protective glove having a radiation absorbing capacity equivalent to that of at least 0.13 mm thickness of lead which comprises forming the glove from a polymeric composition comprising a flexible polymer and containing at least 25% by volume of particulative tungsten material. The gloves of the invention may be formed by any convenient moulding or fabrication process. The gloves may be produced by a process which comprises forming one or more flexible sheets of synthetic polymer containing at least 25% by volume of particulate tungsten material, cutting one or more shaped glove pieces from the sheet or sheets and joining he glove shaped glove piece or pieces at the peripheral edges or margins thereof to form a glove. Suitable shaped glove pieces include a foldable piece having the outline of two opposed glove halves joined for example at the base wrist portions thereof, two glove shaped opposed halves, and individual portions of these shaped pieces. The cutting of the sheet or sheets can conveniently be carried out by a stamping method using shaped dies. The sheet or sheets of tungsten filled polymer can be formed by mixing the appropriate amounts of polymer, tungsten powder and optionally plasticiser and/or filler into the polymer in a conventional rubber mixer such as a heated rubber planetary or Banbury mixer or on a rubber two roll mill and then extruding, casting or calendering the polymer mixture at a suitable temperature onto a cooled smooth surface or substrate. A sheet containing a cross-linking agent of the polymer may be post cured by a suitable heating means. The sheet so formed can conveniently be a continuous sheet from which the shaped glove pieces can be cut. The glove pieces can be joined by a conventional heat-sealing or adhesive process. The glove may be produced by a process which comprises moulding a flexible polymer containing at least 25% by volume of particulate tungsten material. The polymer, filler and optionally plasticiser mixture can be formed by the processes hereinbefore described. Suitable moulding processes include processes in which a glove former is sprayed with or dipped into a solution hot melt or powder suspension of the polymer mixture, processes which comprises injection-moulding compression mouldihg or thermo-forming a melt or plastic mass of the polymer mixture and processes which comprise forming, for example, vacuum forming a sheet of the polymer mixture in a heated mould. Such moulding processes may advantageously provide seamless gloves. The glove may be cured during moulding for example during a reaction injection moulding or after moulding. Gloves of the invention are suitable for surgical or medical use. The gloves can be made to provide a radiation absorption equivalent to that of standard thickness of lead typically 0.13 mm, 0.25 mm, 0.35 mm or 0.5 mm thickness of lead. The thinner wall gloves, which meet one of the two lower standards of lead equivalent radiation protection, will normally be suitable for surgical use. The thicker wall gloves, which meet one of the two higher standards of lead equivalent radiation protection will normally be suitable for medical diagnostic use. Such thicker wall gloves which may e in the form of a gauntlet may also be suitable for non-medical uses for example in the nuclear field. All these gloves of the invention, however, will advantageously provide greater radiation protection than that the gloves disclosed in U.S. Pat. No. 5,001,354.
	description	The present invention relates to a decontamination method and system for soil and the like for decontaminating soil of, for example, fields, and water contaminated with radioactive materials reliably and rapidly on site, aiming to perform decontamination with precision and improved efficiency. The Great East Japan Earthquake occurred on March 2011 caused an accident of the Fukushima Daiichi Nuclear Power Plant of Tokyo Electric Power Company. The accident has dispersed harmful radioactive materials in a wide area and contaminated cities and towns, fields, mountains and forests, the sea, lakes and marshes, and rivers. The radioactive materials adhered to or deposited on persons, animals, and plants have endangered their lives and caused serious damage, stopping various industrial activities such as farming, forestry, livestock farming, fishery, and the like. Removal of such radioactive materials from living environments and industrial activity areas is mandatory for recovery and restart of industrial activities. Decontamination of soil in fields is an urgent problem especially for people engaged in farming. However, contamination has widely spread over fields and scattered throughout villages and mountains as well as plains. Decontaminating such wide areas with human power requires a great amount of time and labor, which is inefficient. The advanced age of many people engaged in farming also makes the decontamination work extremely difficult. In order to deal with the contamination treatment or decontamination treatment of such soil, there is a method for decontaminating halide radioactive wastes by dissolving radioactive material into solvent and then separating the radioactive material from the solvent. In this case, halide is dissolved into water, which is a solvent, to precipitate rare earth elements in the solvent, and the precipitate is collected. As a means for separating non-radioactive material from the solvent, the solvent is evaporated or cooled to precipitate the non-radioactive material (refer, for example, to Patent Document 1). However, in the above decontamination method, the technique in which the radioactive material is collected by dissolving halide in water shows a low collection rate. Further, another technique which includes steps of evaporation and cooling of the solvent requires a heating device and cooling facilities, thereby making the facilities large-scale and expensive. Still another example of soil decontamination methods includes digging the soil contaminated with harmful chemical materials, putting the soil into a hopper of heating device, and heating the soil while washing by using nitrogen to desorb and separate the contaminants in the soil (refer, for example, to Patent Document 2). Such a decontamination method also has problems. The method requires time and labor for moving the contaminated soil to a remote decontamination device and additionally for returning the decontaminated soil to the original position. Further, the contaminated soil needs to be excavated deeply, not only the surface soil. Thus, the method requires an appropriate excavating facility, making the decontamination expensive and large-scale. Additionally, the decontamination device further requires a nitrogen washing machine, heating device, and separator, and thus making the decontamination large scale and expensive. There is still another example of decontamination methods for soil contaminated with radioactive cesium. The contaminated soil is stored in a water supply tank, and carbon-dioxide gas of high partial pressure is injected into the tank to supply hydrogen ions. After extracting cesium ions on the surfaces of soil particles into a liquid phase, the solution is shifted to a separation vessel which is open to air and the carbon-dioxide gas is released into the air. Then, the pH value in the liquid phase is raised to separate therefrom ions such as alkaline earth metals other than cesium and the ions are deposited on carbonates or hydroxides, and the cesium remaining in the liquid phase is condensed and separated (refer, for example, to Non-Patent Document 1). The above-described soil decontamination method also has shortcomings. In the above decontamination method, the liquid phase of supernatant fluid in the water supply tank is sent to the separation vessel so that the supernatant fluid does not contain much cesium, which has a high specific gravity. This leads to a low efficiency of concentration and separation of cesium. Further, cesium accumulates in the lower part of the water supply tank and promotes attachment and deposition on the soil, and thereby lowering the effect of decontamination. Accordingly, use of decontaminated soil has been difficult and impractical. Another decontamination method for soil contaminated with radioactive cesium includes adding water to the contaminated soil received in the reaction vessel, placing positive and negative electrodes in the reaction vessel, applying a voltage to the electrodes to attract radioactive cesium ions onto the negative electrode, depositing soil and other matter associated therewith on the positive electrode, and separating and collecting the radioactive cesium from the contaminated soil to significantly reduce the volume of the contaminants (refer, for example, to Non-Patent Document 2). However, such a decontamination method for soil also has problems. Since the soil is received in the reaction vessel together with other matters, the method requires a high voltage application, which results in poor electrolysis efficiency. The radioactive cesium ions attracted onto the cathode contain foreign matters so that the cesium ions are separated with a low accuracy. Further, since the decontaminated soil also contains other matters, the separation process is time-consuming and the decontaminated soil cannot be used immediately. To solve such problems, the applicants have developed and proposed a decontamination method and system for soil and the like (refer, for example, to Patent Document 3). The decontamination method and system for soil and the like include the steps of: introducing and dissolving an object to be decontaminated contaminated with radioactive materials into an acid eluting solvent, condensing and separating the radioactive materials from the eluting solvent, wherein the object to be decontaminated includes contaminated soil and contaminated water, one or both of which are collected and introduced into the eluting solvent to perform solid-liquid separation of the radioactive materials and the object to be decontaminated dissolved in the eluting solvent, the soil which has been separated from the eluting solvent is further separated into solid and liquid and collected, aiming to reduce the volume of the contaminated soil and to reutilize the soil which includes no radioactive material and to restart of farming, wherein the eluting solvent in which the radioactive materials after solid-liquid separation are dissolved is electrolyzed and condensed, the radioactive cesium ions deposited on an electrode are adsorbed on an adsorbent and collected, the adsorbent on which the radioactive cesium ions are adsorbed is sealed and stored in the container, and the container is stored in storage facilities as needed, thereby achieving a safe disposal of the radioactive materials. The applicants had intended to remove radioactive cesium which has a great impact by the proposed decontamination method for soil and the like, and had not intended to decontaminate radioactive materials other than radioactive cesium, for example, tritium (3H) and tritiated water (HTO6). The tritium, or tritiated hydrogen, is a radioactive isotope of hydrogen with the mass number of three. Tritium is combined with oxygen and exists in water as tritiated water (HTO6). Tritiated water exists in hydrosphere in states of gas, liquid, and solid phases. Tritiated water has been spread widely in vapor, precipitation, groundwater, rivers, lakes, sea water, drink water, and living things. Generally, tritium has been considered as one of the least toxic radionuclides and taken lightly from a standpoint of effects on living things since tritium taken is distributed uniformly in a body, and tritium has a relatively short biological half-life (2.3 years) and has low energy. Tritium emits low beta rays but the radiation reaches only 1 μm in cells. While circulating through the entire body as blood, the tritium hardly attacks gene DNAs. When tritium is taken into a cell nucleus, the distance between the tritium and DNA becomes closer so that tritium starts attacking DNAs as radioactive cesium does. Large amounts of hydrogen exits in DNA. Since tritium has the same chemical properties as those of hydrogen, tritium acts normally if replaced by hydrogen. However, if tritium is changed into helium (He) after emitting radiation, DNAs in the portion where tritium is changed into helium is damaged and then the damage can become a risk, increasing cancer incidence. Such problems have been pointed out. The background art referred to hereinabove is: [Patent Document 1] JP-A-10-213697 [Patent Document 2] JP-A-5-192648 [Patent Document 3] JP-A-2014-41066 [Non-Patent Document 1] Choji, Tetsuji Takada, Eiji Tafu, Masamoto (Toyama National College of Technology) Hara, Masanori (University of Toyama) Houshasei-seshiumu Osendojyo wo Tansan-gasu nomide Senjyo, Shufuku suru Anzen Anshin na Kahan-gata Souchi no Kouchiku (Construction of Safe and Portable Apparatus for Cleaning and Repairing Soil Contaminated with Radioactive Cesium by Using Only Carbon Dioxide Gas),The First Fukushima Conference's SummaryThe Society for Remediation of Radioactive Contamination in Environment, 21 [Non-Patent Document 2] Ueda, Yuko Watanabe, Isao Toida, Hideki Honda, Katsuhisa (Center of Advanced Technology for the Environment, Faculty of Agriculture, Ehime University) Denki Bunkai wo Riyou shita Houshaseibushitsu Jyosen Gijyutsu no Teian (Proposal for Decontamination Technique of Radioactive Materials Using Electrolysis)The First Fukushima Conference's SummaryThe Society for Remediation of Radioactive Contamination in Environment, 92 The object of the present invention is to address such problems and provide a decontamination method and system for soil and the like for decontaminating soil of, for example, fields and water contaminated with radioactive materials reliably and rapidly on site, aiming to perform decontamination with precision and improved efficiency. The invention improves the decontaminated soil by adding soil activators to the decontaminated soil, and then the improved soil is returned to the original field readily to promote the restart of farming. Further, the invention separates the radioactive materials adhering to or deposited on the soil from the soil precisely and then the radioactive materials are condensed. The invention enables a reduction in the volume of the contaminated soil and achieves a safe treatment of the radioactive materials. Further, the invention enables decontamination of radioactive cesium and tritium, thereby relieving anxiety of internal exposure to the radiation of tritium, and enables an efficient and safe disposal of the decontamination apparatus. According to a first aspect of the invention, a decontamination method includes: introducing and dissolving an object to be decontaminated, contaminated with radioactive materials, into an eluting solvent to separate the radioactive materials by elution thereof into the eluting solvent, the object to be decontaminated including contaminated soil and contaminated water and one or both of which being collected and introduced into the eluting solvent; separating the radioactive materials and the object to be decontaminated dissolved in the eluting solvent into solid and liquid; collecting the soil after solid-liquid separation and from which the radioactive materials are removed; electrolyzing a separated liquid containing the eluting solvent and the contaminated water after solid-liquid separation by introducing the separated liquid into an electrolysis tank; depositing metal ions such as the radioactive materials on a cathode; collecting hydrogen containing tritium generated by the electrolysis in the electrolysis tank; and trapping the hydrogen moved to an outside of the electrolysis tank. Accordingly, decontamination of tritium contained in the hydrogen is achieved, thereby relieving anxiety of internal exposure to the radiation of tritium. The eluting solvent typically comprises water. Soil, as a whole, does not dissolve in water, so the herein use of the term “dissolved” regarding soil in the eluting solvent is intended to encompass “dispersed”. According to a second aspect of the invention, the electrolysis is performed in the electrolysis tank, and the electrolysis tank is hermetically sealed. Accordingly, hydrogen generated by the electrolysis can reliably be collected. According to a third aspect of the invention, oxygen is discharged from the electrolysis tank upon a pressure of the oxygen accumulated in the electrolysis tank reaches a predetermined value or more during electrolysis. Accordingly, safety during electrolysis can be ensured. A fourth aspect of the invention is one or a plurality of deposition members, to which an electric current is supplied from the cathode, being arranged around the cathode, and the metal ions such as the radioactive materials are deposited on the cathode and the deposition members. Accordingly, radioactive cesium, other radioactive materials, and the metal ions such as heavy metal in the separated liquid are removed precisely and decontaminated. According to a fifth aspect of the invention, the separated liquid is introduced into an adsorption filter before being introduced to the electrolysis tank, and the metal ions such as the radioactive materials are adsorbed by the adsorption filter. Accordingly, the metal ions such as the radioactive materials are efficiently adsorbed on the adsorption filter and in the electrolysis tank separately, thereby improving the time for electrolysis in the electrolysis tank. According to a sixth aspect of the invention, the hydrogen is charged into a gas cylinder for storage. Accordingly, tritium contained in hydrogen is prevented from leaking and stored safely. According to a seventh aspect of the invention, the cathode, the deposition member, the metal ions such as the radioactive materials deposited on the cathode and the deposition member, a collector for collecting hydrogen, remaining separated liquid, and the adsorption filter are left as they are in the used electrolysis tank and stored. Accordingly, inconvenience of storing them separately can be avoided, and thus, efficient and safe storage can be achieved easily and at low cost. According to an eighth aspect of the invention, tritiated water is mixed in the contaminated water. Accordingly, tritium contained in the tritiated water is removed and decontaminated by collecting hydrogen generated by electrolyzing the contaminated water and the separated liquid that contain tritiated water. According to a ninth aspect of the invention, the radioactive materials include radioactive cesium and tritium. Accordingly, the objects to be decontaminated include radioactive cesium, which is highly reactive and has a relatively long-term half-life, and tritium, which has a relatively short half-life and has been considered as one of the least toxic radionuclide and thus taken lightly. Accordingly, decontamination of various types of radioactive materials is achieved. According to a tenth aspect of the invention, a decontamination system for soil and the like includes: a separation vessel for receiving an object to be decontaminated, contaminated with radioactive materials, and an eluting solvent, and for dissolving the radioactive materials into the eluting solvent; a solid-liquid separation filter for separating the radioactive materials dissolved into the eluting solvent and the object to be decontaminated into solid and liquid; an electrolysis tank for receiving and electrolyzing a separated liquid including the eluting solvent and contaminated water after solid-liquid separation; the object to be decontaminated including contaminated soil and contaminated water, the soil after solid-liquid separation and from which the radioactive materials are removed being collected, metal ions such as the radioactive materials being deposited on a cathode; a collector for collecting hydrogen provided in the electrolysis tank and collects hydrogen containing tritium generated by electrolysis, whereby the hydrogen is moved to an outside of the electrolysis tank and trapped. Accordingly, decontamination of tritium contained in the hydrogen is achieved, thereby relieving anxiety of internal exposure to the radiation of tritium. According to an eleventh aspect of the invention, the electrolysis tank includes a hollow airtight container, the cathode is provided at the center portion of the airtight container, and the airtight container is set as an anode and a positive electric potential is applied. Accordingly, electrolysis can be performed in the airtight container, and the generated hydrogen is prevented from leaking, thereby performing electrolysis safely. A twelfth aspect of the invention is a safety valve provided in a space above the collector in the electrolysis tank, and oxygen is discharged from the electrolysis tank when a pressure of the oxygen accumulated in the electrolysis tank reaches a predetermined value or more. Accordingly, safety during electrolysis can be ensured. A thirteenth aspect of the invention is one or a plurality of deposition members, wherein the deposition member is provided below the collector to surround the cathode, and the metal ions such as the radioactive materials are deposited on the cathode and the deposition members. Accordingly, metal ions such as the radioactive materials are deposited precisely and efficiently. According to a fourteenth aspect of the invention, one end of a trap conduit is provided in the collector and the other end is provided at a hydrogen-gas charging device, the hydrogen-gas charging device includes a suction pump for suctioning hydrogen in the collector, and a charging device for injecting the hydrogen into a gas cylinder. Accordingly, hydrogen is suctioned from the collector and charged into the gas cylinder safely. A fifteenth aspect of the invention is an adsorption filter for adsorbing the metal ions such as the radioactive materials in the separated liquid, and the adsorption filter is provided in the electrolysis tank. Accordingly, the filter adsorbs the metal ions such as the radioactive materials rationally, together with the electrolysis tank, and the used electrolysis tank is stored easily. According to a sixteenth aspect of the invention, the airtight container has a bottom coated with insulating materials. Accordingly, generation of oxygen at the inside bottom of the airtight container is prevented, and thus hydrogen can be collected safely. A seventeenth aspect of the invention is a decontamination vehicle which can be moved to a collection site of the object to be decontaminated, wherein the vehicle is equipped with the separation vessel, the electrolysis tank, the solid-liquid separation filter, a carbon-dioxide gas cylinder for generating an eluting solvent, a water supply tank, and a hydrogen-gas charging device. Accordingly, the radioactive materials including tritium is decontaminated at the collection site of the object to be decontaminated organically and rationally. According to a first aspect of the invention, hydrogen containing tritium generated by electrolysis is collected in the electrolysis tank and the hydrogen is moved to an outside of the electrolysis tank and trapped. Accordingly, decontamination of tritium contained in the hydrogen is achieved, thereby relieving anxiety of internal exposure to the radiation of tritium. According to a second aspect of the invention, the electrolysis is performed in the electrolysis tank, and the electrolysis tank is hermetically sealed. Accordingly, hydrogen generated by the electrolysis can be reliably collected. According to a third aspect of the invention, oxygen is discharged from the electrolysis tank when a pressure of the oxygen accumulated in the electrolysis tank reaches a predetermined value or more during electrolysis. Accordingly, safety during electrolysis can be ensured. A fourth aspect of the invention is one or a plurality of deposition members, to which an electric current is supplied from the cathode, being arranged around the cathode, and the metal ions such as radioactive materials are deposited on the cathode and the deposition members. Accordingly, radioactive cesium, other radioactive materials, and the metal ions such as heavy metal in the separated liquid are removed precisely and decontaminated. According to a fifth aspect of the invention, the separated liquid is introduced into an adsorption filter before being introduced to the electrolysis tank, and the metal ions such as the radioactive materials are adsorbed by the adsorption filter. Accordingly, the metal ions such as radioactive materials are efficiently adsorbed on the adsorption filter and in the electrolysis tank separately, thereby improving the time for electrolysis in the electrolysis tank. According to a sixth aspect of the invention, the hydrogen is charged into a gas cylinder for storage. Accordingly, tritium contained in hydrogen can be prevented from leaking and stored safely. According to a seventh aspect of the invention, the cathode, the deposition member, metal ions such as the radioactive materials deposited on the cathode and the deposition member, a collector for collecting hydrogen, remaining separated liquid, and the adsorption filter are left as they are in the used electrolysis tank and stored. Accordingly, inconvenience of storing them separately can be avoided, and thus, efficient and safe storage can be achieved easily and at low cost. According to an eighth aspect of the invention, tritiated water is mixed in the contaminated water. Accordingly, tritium contained in the tritiated water can be removed and decontaminated by collecting hydrogen generated by electrolyzing the contaminated water and the separated liquid that contain tritiated water. According to a ninth aspect of the invention, the radioactive materials include radioactive cesium and tritium. Accordingly, the objects to be decontaminated include radioactive cesium, which is highly reactive and has a relatively long-term half-life, and tritium, which has a relatively short half-life and has been considered as one of the least toxic radionuclides and thus taken lightly. Accordingly, decontamination of various types of radioactive materials can be achieved. According to a tenth aspect of the invention, a collector for collecting hydrogen is provided in the electrolysis tank, the hydrogen containing tritium generated in electrolysis is collected, and the hydrogen is moved to the outside of the electrolysis tank and trapped. Accordingly, decontamination of tritium contained in the hydrogen is achieved, thereby relieving anxiety of internal exposure to the radiation of tritium. According to an eleventh aspect of the invention, the electrolysis tank includes a hollow airtight container, the cathode is provided at the center portion of the airtight container, and the airtight container is set as an anode and a positive electric potential is applied. Accordingly, electrolysis can be performed in the airtight container, and the generated hydrogen is prevented from leaking, thereby performing electrolysis safely. A twelfth aspect of the invention is a safety valve provided in a space above the collector in the electrolysis tank, and oxygen is discharged from the electrolysis tank upon a pressure of the oxygen accumulated in the electrolysis tank reaches a predetermined value or more. Accordingly, safety during electrolysis can be ensured. A thirteenth aspect of the invention is one or a plurality of deposition members, wherein the deposition member is provided below the collector to surround the cathode, and the metal ions such as the radioactive materials are deposited on the cathode and the deposition members. Accordingly, the metal ions such as the radioactive materials can be deposited precisely and efficiently. According to a fourteenth aspect of the invention, one end of a trap conduit is provided in the collector and the other end is provided at a hydrogen-gas charging device, the hydrogen-gas charging device includes a suction pump for suctioning hydrogen in the collector, and a charging device for injecting the hydrogen into a gas cylinder. Accordingly, hydrogen is suctioned from the collector and can be charged into the gas cylinder safely. A fifteenth aspect of the invention is an adsorption filter for adsorbing the metal ions such as the radioactive materials in the separated liquid, and the adsorption filter is provided in the electrolysis tank. Accordingly, the filter adsorbs the metal ions such as the radioactive materials rationally, together with the electrolysis tank, and the used electrolysis tank is stored easily. According to a sixteenth aspect of the invention, the airtight container has a bottom coated with insulating materials. Accordingly, generation of oxygen at the inside bottom of the airtight container is prevented, and thus hydrogen can be collected safely. A seventeenth aspect of the invention is a decontamination vehicle which can be moved to a collection site of the object to be decontaminated, wherein the vehicle is equipped with the separation vessel, the electrolysis tank, the solid-liquid separation filter, a carbon-dioxide gas cylinder for generating an eluting solvent, a water supply tank, and a hydrogen-gas charging device. Accordingly, the radioactive materials including tritium can be decontaminated at the collection site of the object to be treated organically and rationally. Embodiments of the present invention will be described below with reference to the drawings, in which the invention is applied to decontamination of soil and contaminated water in fields, paddy fields, wetlands, and the like. In FIGS. 1 through 13, reference numeral 1 is an object area to be decontaminated and is contaminated with radioactive materials. The object area includes a field 3 with contaminated surface soil 2 containing a predetermined moisture or with contaminated surface soil 2 which is dried and hardened. The object area also includes a paddy field 5, the surface soil 2 of which is submerged into the contaminated water 4, and wetlands having a large amount of water. The invention can deal with decontamination for both cases. In drawings, reference numeral 6 represents weeds on the surface soil 2 of the field 3, and reference numeral 7 represents rice plants and weeds on the surface soil 2 in the contaminated water 4. A decontamination vehicle 9 is parked in a farm road 8 or a space adjacent to the object area to be decontaminated 1. A suction hose 11 is pulled out from a decontamination tank 10, and a predetermined contaminated soil or contaminated water 4 which includes tritiated water (HTO6) is sucked and collected from the end of the hose 11. In drawings, reference numeral “d” represents a depth of suction or collection of the surface soil 2 in the field 3 and the depth corresponds to a depth of penetrated cesium, which is a radioactive material. In some embodiments, 5 cm or more of the surface soil 2 is collected. The decontamination vehicle 9 is configured by modifying a conventional vacuum truck. The vehicle body is equipped with a separation vessel 10, a water supply tank 12, an electrolysis tank 13, a gas cylinder 14 charged with carbon dioxide to a predetermined pressure, and a suction pump 15 for suctioning a solid, liquid, and gas. The separation tank 10 is formed by a box container having an openable lid. The separation tank 10 is provided at the top of the lid with a rotatable cylindrical reel 16 around which the suction hose 11 can be wound. The reel 16 is rotatably biased in the counterclockwise direction in FIG. 3 via a recoil spring (not shown) and around which the suction hose 11 can be wound. When the contaminated soil 17 is sucked, the suction hose 11 is pulled outwardly. The pulling force enables the reel 16 to rotate in the clockwise direction in FIG. 3 and thereby unreeling the suction hose 11. One end of the suction hose 11 is in communication with the inside of the separation vessel 10. The contaminated soil 17 or contaminated water 4 sucked from the tip end is introduced into the separation vessel 10. The base part of the suction hose 11 is provided with an on-off valve 18. The other end of the suction hose 11 is provided with a filter 19 for preventing foreign matter to be sucked. In drawings, reference numeral 20 is a cylindrical hose guide concentrically provided outside the reel 16, and the tangent portion of the hose guide is formed with a hose insertion hole 21. Reference numeral 22 is a hose clamp provided at the rear end of the separation vessel 10, and reference numeral 23 is a bottomed hose receptacle. The water supply tank 12 is formed by a box container having an openable lid and is provided adjacent to the separation vessel 10. The water supply tank 12 stores clean water 24 therein for supplying a predetermined amount of water to the separation vessel 10 and the electrolysis tank 13. In this case, it is preferable that a heater be provided on the periphery of the water supply tank 12 to prevent the water 24 from freezing. The on-off valves 25, 26 are provided at the bottom of the water supply tank 12. Respective one ends of water supply conduits 27, 28 are connected to the on-off valves 25, 26. The other end of the water supply conduit 27 is provided at the upper part in the separation vessel 10, and the other end of the water supply conduit 28 is connected to an on-off valve 29 provided at the bottom of the electrolysis tank 13. The electrolysis tank 13 includes a cylindrical airtight container 30 made of stainless steel plate, and it has a capacity of about 1.8 L. The surface of the electrolysis tank 13 is coated with lead to shield radiation. The electrolysis tank 13 is provided isolated from the vehicle body and the adjacent parts. The airtight container 30 has a bottom surface with an insulating coating 30a to prevent generation of oxygen from the coated portion. The oxygen is generated from the inner surface of the side wall except the portion with the insulating coating 30a. The generated oxygen moves up along the inner surface of the side wall and accumulates in the upper space of a collector 35 provided at the top of the airtight container 30. It is preferable that a heater (not shown) be provided on the outer periphery of the electrolysis tank 13 for heating to promote electrophoresis of, for example, cesium ions and heavy metal ions. A rod-like cathode 31 is vertically provided by passing through the center of the airtight container 30. A lead 32 is wired to the cathode 31 and the airtight container 30, which is an anode, and the lead 32 is connected to a DC power 33 and a switch 34. The lower end of the cathode 31 is provided immediately above the bottom of the airtight container 30. A cylindrical collector 35 for collecting hydrogen is provided on the middle and high portion of the cathode 31 in the airtight container 30. The collector 35 is a deep cylindrical container having an opening on one side and arranged with the opening faces downwardly. One or a plurality of deposition members 36 is provided adjacent to each other at the lower part of the inner side of the collector 35. The deposition member 36 is placed to surround the cathode 31 and electrically connected to the cathode 31. The deposition member 36 according to the present embodiment is made of a wire-netting drum, a metal plate, a metal rod, and the like. The deposition member 36 is submerged in a separated liquid 37 introduced from a solid-liquid separation filter, described later, during electrolysis. The gas cylinder 14 stands in a space defined by the water supply tank 12 and the electrolysis tank 13. An on-off valve 44 is provided at the top end of the gas cylinder 14. The on-off valve 44 is connected to one end of a gas conduit 45, and the other end of the gas conduit 45 is connected to an on-off valve 46 provided at the bottom of the separation vessel 10. The three-way valve 47 is provided at an upper stream of the gas conduits 45. One end of the gas conduit 48 is connected to the three-way valve 47, and the other end of the gas conduit 48 is connected to the lower peripheral surface of the electrolysis tank 13 to supply carbon dioxide. A trap 38 has one end provided above the liquid surface of the separated liquid 37 and at the upper part of the collector 35. The other end of the trap 38 passes through the collector 35 and the airtight container 30 and connects to a hydrogen-gas charging device 39 provided outside. The hydrogen-gas charging device 39 includes a suction pump 40 and a gas cylinder 41 for charging hydrogen gas. The inlet of the gas cylinder 41 is provided removably with the end of the trap 38 and with an on-off valve (not shown) which is normally closed. In drawings, reference numeral 42 denotes an adsorption filter such as zeolite provided at the top of the airtight container 30 and the filter adsorbs the radioactive materials and heavy metals in the separated liquid 37 introduced from a solid-liquid separation filter, which will be described later. Reference numeral 43 represents a pH sensor provided in the separated liquid 37 in the airtight container 30 for measuring acid concentration of the separated liquid 37. After water is supplied to the separation vessel 10, carbon dioxide is supplied to the separation vessel 10 from the gas cylinder 14 via the gas conduit 45. Carbonated water (H2CO3) 49 having a predetermined acid concentration is prepared by using the carbon dioxide in the separation vessel 10 as an eluting solvent for radioactive cesium. In some embodiment, the acid concentration of the carbonated water 49 in the separation vessel 10 is set to pH 3 to 7 and the carbonated water 49 is used as an electrolyte in the electrolysis tank 13. In drawings, reference numeral 50 denotes a stirring fan provided at the bottom of the separation vessel 10. An on-off valve 51 is provided at the bottom of the separation vessel 10. A solid-liquid conduit 52 has one end connected to the on-off valve 51, and the other end connected to a tubular solid-liquid separation filter 53 which is vertically long. The solid-liquid separation filter 53 is vertically provided and a rotating tube (not shown) is provided therein. A centrifuge (not shown) is provided inside the rotating tube. In the solid-liquid components introduced from the separation vessel 10, the heavy soil 17 is moved to the outer side of the rotating tube and the carbonated water 49 which includes a mixture of radioactive cesium ions and tritiated water is move to the inner side of the rotating tube by their specific gravities to perform solid-liquid separation. The soil 17a which has been separated from the radioactive material and the carbonated water is allowed to sink and accumulate in the lower part of the separation filter 53 so that the soil 17a can be collected from outside and the separated liquid 37 containing radioactive cesium, strontium, and heavy metals can be sent to the electrolysis tank 13. A discharge pipe 54 is protruded from the lower part of the solid-liquid separation filter 53. The discharge pipe 54 is provided with an openable and closable discharge valve 55. The soil 17a can be collected by opening the discharge valve 55. Accordingly, if a plurality of solid-liquid separation filters 53 is provided in the solid-liquid conduit 52, the carbonated water 49 containing the soil 17a, radioactive cesium ions, and tritiated water can be separated with high accuracy and efficiently. A separation-liquid duct 56 has one end connected to the center of the top end of the solid-liquid separation filter 53 and the other end connected the adsorption filter 42 via the on-off valve 57, and the end of the duct is placed at the bottom of the airtight container 13. Then, radioactive cesium, strontium, and the metal ions such as heavy metals are deposited on the cathode 31 and plural deposition members 36 in the electrolysis tank 13. After collecting hydrogen in the collector 35 generated by electrolysis, the centrifuge is stopped and then clean carbonated water 49 which includes no radioactive material in the electrolysis tank 13 is returned to the separation vessel 10 via a return pipe (not shown). A pair of loop conduits 58, 59 has respective one ends connected to the suction pump 15 and respective other ends connected to the four-way valve 60. A vent pipe 61 having one end open to the atmosphere and a communication pipe 62 having one end provided in the separation vessel 10 are connected to the rest of the two ports of the four-way valve 60. The piping port of the four-way valve 60 can be switched by a switching lever 63 and the switching position thereof includes a neutral position, discharge position, and suction position, and is normally set at the neutral position. Accordingly, communication between the vent pipe 61 and the communication pipe 62, both connected to the four-way valve 60, and suction and discharge operation of the suction hose 11 provided in the separation vessel 10 can be controlled. When the contaminated soil 17 is collected and introduced into the separation vessel 10, the suction pump 15 is actuated, and the switching lever 63 is switched from the neutral position to the suction position, as shown in FIG. 6. The loop conduits 58, 59 are allowed to communicate with the communication pipe 62 to create a negative pressure in the separation vessel 10. The contaminated soil 17 is suctioned from the end portion of the suction hose 11 and introduced into the separation vessel 10 and immersed into the carbonated water 49. The contaminated soil 17 introduced into the separation vessel 10 is then subjected to diproton acid cleaning by the carbonated water 49. Thus, the metal ions such as radioactive cesium ions, strontium, and heavy metals are ionized and conveyed to the solid-liquid separation filter 53 and electrolysis tank 13 together with the contaminated soil 17 and the carbonated water 49. At that time, the suction pump 15 is actuated and the switching lever 63 is switched from the suction position to the discharge position, as show in FIG. 7. The loop conduits 58, 59 are allowed to communicate with the vent pipe 61 and the communication pipe 62, and the on-off valves 18, 46 are closed. The on-off valves 51, 57 are opened to suction air from the vent pipe 61 and the air is sent to the communication pipe 62 from the loop conduits 58, 59 to pressurize the separation vessel 10. A muddy-liquid mixture including the soil 17, the metal ions such as radioactive cesium ions, strontium, heavy metals, and carbonated water 49, that have been separated, in the separation vessel 10 is sent to the solid-liquid conduit 52, and then to the solid-liquid separation filter 53. The muddy-liquid mixture is separated into solid and liquid, and the separated liquid 37 including the radioactive materials, the metal ions such as heavy metals, and the carbonated water 49, that have been separated, is sent to the electrolysis tank 13. Then, the separated liquid 37 is introduced to the adsorption filter 42 before being introduced to the electrolysis tank 13. The filter 42 adsorbs the radioactive materials and the metal ions such as the heavy metals in the separated liquid 37. The filtered separated liquid 37 that has passed through the filter 42 is introduced into the electrolysis tank 13 for electrolysis. Electrolysis in the electrolysis tank 13 is as shown in FIG. 8. A switch 34 is tuned on to pass a current between the cathode 31 and the airtight container 30, which is an anode. Hydrogen is generated on the cathode 31, and oxygen is generated in the airtight container 30 which is an anode. A trace amount of tritium exists in the hydrogen, and the bubbles of the hydrogen containing tritium move up along the cathode 31 and the deposition member 36. When the bubbles reach the liquid level of the separated liquid 37, the bubbles are collected in the collector 35. Then, the suction pump 40 is actuated to suction hydrogen containing the collected tritium. The hydrogen is introduced into the trap 38 and injected into the gas cylinder 41. The hydrogen is charged into the gas cylinder 41 to about the atmospheric pressure. Oxygen is generated from the side surfaces of the airtight container 30. The air bubbles move up along the side surfaces and to the upper space in the collector 35 provided at the upper part of the airtight container 30, whereby the bubbles of oxygen accumulate in the space. In this case, the insulating coating 30a is formed on the bottom of the airtight container 30 so that oxygen is not generated from the insulated portion. A safety valve 72, which is normally closed, is attached to the top of the airtight container 30. When the accumulated oxygen in the airtight container 30 reaches a predetermined pressure or more, the safety valve 72 is opened to discharge the oxygen to the outside via a discharge pipe 73. The remained radioactive materials and the metal ions such as heavy metals filtered through the adsorption filter 42 during electrolysis are electrophoresed in the separated liquid 37. Then, they are deposited on the cathode 31 and the deposition members 36, to which the same potential as the cathode 31 is applied. As described above, the separated liquid 37 in the airtight container 30 is electrolyzed to deposit the radioactive materials and the metal ions such as heavy metals on the cathode 31 and the deposition member 36, while hydrogen containing tritium is charged into the gas cylinder 41 to remove the metal ions containing the radioactive materials and hydrogen gas from the separated liquid 37 and to prepare clean water containing the carbonated water 49. The separated liquid 37 containing the carbonated water 49 which has been cleaned in the manner as described above is returned to the separation vessel 10 and utilized. In that case, the suction pump 15 is actuated as shown in FIG. 10. The switching lever 63 is switched from the discharge position to the suction position to close the on-off valves 18, 46. The on-off valves 51, 57 are opened to suck muddy water containing the carbonated water 49 in the separated liquid 10 via the communication pipe 62, thereby creating a negative pressure in the separation vessel 10. The clean separated liquid 37 in the separation vessel 13 is suctioned with the separation-liquid duct 56 and then introduced to the separation vessel 10 through the solid-liquid conduit 52. The decontaminated soil 17a that has been collected is dried by the sun or by heating. After the decontaminated soil 17 is dried, a predetermined amount of soil activator 65 is added and mixed to decontaminate and modify the collected soil 17. The soil activator 65 includes organic fertilizers such as compost, mycorrhizal fungi, or various kinds of chemical fertilizers including nitrogen, phosphorus, and potassium. The soil 17a modified is returned to the original field in which the soil has been collected, thereby reducing the volume of the contaminated soil 17. Reference numeral 64 represents a pH sensor for measuring acid concentration of the carbonated water 49 in the separation vessel 10. Reference numeral 66 denotes a decontamination worker. In some embodiments, the powder fire extinguishant used as a soil activator 65 may be one for which the expiry date has passed. The powder fire extinguishant contains monobasic ammonium phosphate or ammonium sulfate, whereby the fire extinguishant can be used effectively, despite its expiry date. A predetermined amount of the radioactive materials accumulate on the cathode 31 and deposition member 36 in the electrolysis tank 13 by electrolysis so that the used electrolysis tank 13 needs to be disposed of and replaced based on the electrolysis throughput of the separated liquid 37. When the used electrolysis tank 13 is disposed, the on-off valve 57 is detached from the airtight container 30, and the separation-liquid duct 56 is cut at the middle portion to disconnect the solid-liquid separation filter 53. At that time, the adsorption filter 42 also needs to be disposed of and replaced. However, the adsorption filter 42 is placed in the airtight container 30 and thus can be disposed of and replaced at the same time as the electrolysis tank 13. Further, the lead 32 is cut and reused with the electrode 33 and the switch 34. In addition, the pH sensor 43 can be reused. Further, the trap 38 is cut at the middle portion to reutilize the suction pump 40 and the discharge pipe thereof. The water supply conduit 28 is cut at the end to reutilize the remaining portions of the on-off valves 29, 26 and the water supply conduit 28. As described above, the peripheral components of the used electrolysis tank 13 is removed and formed into a cylindrical shape. The airtight container 30 stores the cathode 31, the plural deposition members 36, the metal ions containing the radioactive materials deposited on these components, the collector 35, the remaining separated liquid 37, the adsorption filter 42, and the safety valve 72, that are left after use. Then the electrolysis tank 13 is stacked in an upright position and then stored in a safe storage facilities 67, which is made of concrete. This state is shown in FIG. 11 and FIG. 12. Meanwhile, after the used electrolysis tank 13 is removed, a new electrolysis tank 13 is insulated and installed at the same position on the decontamination vehicle 9. The separation-liquid duct 56 has one end connected to the solid-liquid separation filter 53 and the other end connected the on-off valve 57. One end of the trap 38 is attached to the inside of the airtight container 30, and the other end is connected to the suction pump 40. A discharge pipe of the suction pump 40 is inserted in a new gas cylinder 41. Additionally, one end of the lead 32 is connected to the top end of the cathode 31, and the other end is connected to the airtight container 30. The pH sensor 43 is attached to the airtight container 30 and replaced. Storage facilities 69, which is similar as the storage facilities 67, is provided for storing the gas cylinder 41 charged with hydrogen gas. The gas cylinder 41 is placed horizontally and stored in the facilities 69 as shown in FIG. 13. In this case, a valve (not shown), which is normally closed, is provided at the mouth portion of the gas cylinder 41 to prevent the charged hydrogen gas from leaking. Additionally, in drawings, reference numerals 70, 71 denote water supply pumps for water supply conduits 27, 28 provided at the lower part of the decontamination vehicle 9. The decontamination method and system for soil and the like configured as described above require a decontamination vehicle 9 configured by modifying a conventional vacuum truck. The vehicle body is equipped with the separation vessel 10, the water supply tank 12, the electrolysis tank 13, the gas cylinder 14 charged with carbon dioxide, the suction pump 15 for suctioning soil, plants, and accumulated water in fields, the solid-liquid separation filter 53, the soil activator 65, and the adsorption filter 42. This state is shown in FIGS. 2 and 3. The separation tank 10 is formed by a box container having an openable lid. The separation tank 10 is provided at the top with a rotatable cylindrical reel 16 around which the suction hose 11 can be wound. The reel 16 is rotatably biased in the counterclockwise direction via a recoil spring (not shown) and around which the suction hose 11 can be wound. When the contaminated soil is suctioned, the suction hose 11 is pulled outwardly and the reel 16 is rotated in the clockwise direction in FIG. 3, thereby unreeling the suction hose 11. The suction hose 11 has one end which is in communication with the inside of the separation vessel 10 and has the other end for introducing the sucked contaminated soil 17 into the separation vessel 10. The base part of the suction hose 11I is provided with an on-off valve 18. The other end of the suction hose 11 is provided with a filter 19 for preventing foreign matter to be sucked. The water supply tank 12 is formed by a box container having an openable lid and placed adjacent to the separation vessel 10. The water supply tank 12 stores a predetermined amount of clean water 24 therein which is to be supplied to the separation vessel 10 and the electrolysis tank 13. The on-off valves 25, 26 are provided at the bottom of the water supply tank 12. Respective one ends of water supply conduits 27, 28 are connected to the on-off valves 25, 26. The other end of the water supply conduit 27 is provided at the upper part in the separation vessel 10, and the other end of the water supply conduit 28 is connected to the on-off valve 29 provided at the bottom of the electrolysis tank 13. The water supply conduits 27, 28 are respectively provided with the water supply pumps 70, 71. The electrolysis tank 13 includes the cylindrical airtight container 30 made of stainless steel plate and has a capacity of about 1.8 L. The surface of the electrolysis tank 13 is coated with lead to shield radiation. The electrolysis tank 13 is provided isolated from the vehicle body and the adjacent components. The electrolysis tank 13 is placed at the front space of a loading platform of the decontamination vehicle 9 and adjacent to the gas cylinder 14 and the water supply tank 12. The rod-like cathode 31 is vertically provided and passes through the center of the airtight container 30. The lead 32 is wired to the cathode 31 and the airtight container 30, which is an anode, and the lead 32 is connected to a DC power 33 and a switch 34. The lower end of the cathode 31 is provided immediately above the bottom of the airtight container 30. A cylindrical collector 35 for collecting hydrogen gas is provided on the middle and high portion of the cathode 31 in the airtight container 30. The collector 35 is formed in a deep cylindrical container having an opening on one side and provided in the airtight container 30 with the opening facing downwardly. The adsorption filter 42 is attached to the top of the airtight container 30, and the filter 42 is connected to the separation-liquid duct 56. The lower end of the separation-liquid duct 56 is placed immediately above the bottom in the airtight container 30, the safety valve 72 is attached to the top of the inside of the airtight container 30, and the discharge pipe 73 is open to the outside of the container 30. Additionally, a plurality of deposition members 36 is provided adjacent to each other and at the inner side of the collector 35. The deposition members 36 are placed to surround the cathode 31 and electrically connected to each other. The deposition members 36 are submerged into the separated liquid 37 during electrolysis. The trap 38 has one end provided above the liquid surface of the separated liquid 37 and at the upper part of the collector 35. The other end of the trap 38 passes through the collector 35 and the airtight container 30 and connects to a hydrogen-gas charging device 39 provided outside. The hydrogen-gas charging device 39 includes a suction pump 40 and a gas cylinder 14 for charging hydrogen gas. The inlet of the gas cylinder 14 is normally closed and provided with an on-off valve (not shown) to which the end of the trap 38 is removably attached. The hydrogen-gas charging device 39 is placed adjacent to the electrolysis tank 13. The gas cylinder 14 stands adjacent to the electrolysis tank 13. An on-off valve 44 is provided at the top end of the gas cylinder 14. The on-off valve 44 is connected to one end of a gas conduit 45, and the other end of the gas conduit 45 is connected to an on-off valve 46 provided at the bottom of the separation vessel 10. The three-way valve 47 is provided in the gas conduits 45. One end of the gas conduit 48 is connected to the three-way valve 47, and the other end of the gas conduit 48 is connected to the lower peripheral surface of the electrolysis tank 13 to supply carbon dioxide to the water supply tank 12 and the electrolysis tank 13 selectively. After water is supplied to the separation vessel 10, carbon dioxide is supplied to the separation vessel 10 from the gas cylinder 14 via the gas conduit 45. Carbonated water (H2CO3) 49 having a predetermined acid concentration is prepared by using the carbon dioxide in the separation vessel 10 as an eluting solvent for radioactive cesium. The on-off valve 51 is provided at the bottom of the separation vessel 10. A solid-liquid conduit 52 has one end connected to the on-off valve 51, and the other end connected to a tubular solid-liquid separation filter 53 which is vertically long. The solid-liquid separation filter 53 having a rotating tube (not shown) therein is placed vertically. The rotating tube includes a centrifuge (not shown) therein. In the solid-liquid components introduced from the separation vessel 10, the soil 17 is moved to the outer side of the rotating tube, while a fluid contaminated lightly including the carbonated water 49 with radioactive cesium ions and tritiated water mixed therein is move to the inner side of the rotating tube to perform solid-liquid separation. The decontamination vehicle 9 includes the separation vessel 10, the water supply tank 12, the small tank 13, the gas cylinder 14, the suction pump 15, the solid-liquid separation filter 49, and the suction hose 11 which can be wound being provided on the top of the separation vessel 10. These components are efficiently and compactly provided. Thus, the decontamination vehicle 9 can be minimized in size and weight, and provided at a lower cost. Further, the decontamination vehicle 9 can be moved to terraced paddy fields in mountains and narrow farm roads in rural areas. The decontamination vehicle 9 can perform a series of decontamination work with the mounted equipment by taking advantage of its mobility, without requiring any heavy machinery for collecting the soil 17. Next, when the soil 17 and the water 4 contaminated with the radioactive materials are decontaminated by using the decontamination vehicle 9, a predetermined amount of clean water 24 is stored in the water supply tank 12 in advance because clean water 24 may not be obtainable on site. Additionally, a predetermined amount of water 24 is stored in the separation vessel 10. The decontamination vehicle 9 is moved to the object area to be decontaminated 1, such as the field 3, the paddy field 5, mountains and forests, fallow farmlands, or lakes and marshes, and parked on, for example, an adjacent farm road 8. This state is shown in FIG. 1. FIG. 5 shows a state of the separation vessel 10, the water supply tank 12, the electrolysis tank 13, and the suction pump 15 before decontamination is started. At the beginning of the decontamination work, the carbon dioxide charged into the gas cylinder 14 is sent to the water 24 in the separation vessel 10 via the gas conduit 45, and the agitator 50 is actuated to agitate the carbon dioxide and the water 24, thereby forming carbonated water 49 having a predetermined acid concentration based on the pH sensor 42. In some embodiments, the acid concentration of the carbonated water 49 is set to pH 3 to 6. In this case, since the carbon dioxide is pressurized to atmospheric pressure or more, and more pressurized carbon dioxide is dissolved into the water 24, thereby facilitating a rise in the acid concentration of the carbonated water 49. In some embodiments, the weakly-acidic carbonated water 49 formed by the carbon dioxide and the water 24 is used as an eluting solvent for cesium. Accordingly, strong acid, such as oxalic acid, which is expensive and hazardous in handling, is not required, and a decontamination work, which will be described later, can be performed safely. After the carbonated water 49 is prepared as described above, the suction hose 11 is unreeled from the separation vessel 10. The worker 66 moves to a predetermined position for the decontamination work, holding the suction hose 11. Before or after unreeling the suction hose 11, the suction pump 15 is actuated and the switching lever 63 is switched to the intake position. The air in the separation vessel 10 is sucked through the vent pipe 61 and the communication pipe 62, thereby enabling the suction hose 11 to suck the object from the end. This state is shown in FIG. 6. In such a situation, the end of the suction hose 11 is positioned immediately above the surface soil 2 of the contaminated field 3. In the case of the paddy field 5 or wetlands, the end of the suction hose 11 is submerged in the contaminated water 4 to suck the contaminated soil 17, contaminated water 4, and tritiated water mixed in the contaminated water 4, that are immediately below the suction hose 11. This state is shown in FIG. 1 and FIGS. 4(a) and 4(b). Then, the contaminated water 4, tritiated water, and contaminated soil 17 are sucked from the tip end of the suction hose 11, and moved to the separation vessel 10 guided by the suction hose 11. This state is shown in FIG. 6. A mixture of the soil 17, the contaminated water 4, and tritiated water suctioned as described are moved to the upper part of the separation vessel 10 guided by the suction hose 11, and fall in the carbonated water 49 in the separation vessel 10 from the open end of the suction hose 11. This state is shown in FIG. 6. Thus, the radioactive cesium ions adhered to or deposited on the soil 17 or the contaminated water 4 are cleaned and dissolved in the carbonated water 49, and the cesium ions that have been separated from the soil 17 and the contaminated water 4 exist in the carbonated water 49. In this case, the acid concentration of the carbonated water 49 gradually decreases as the soil 17 and the contaminated water 4 are introduced. The changes in the acid concentration are checked with the pH sensor 42 and the acid concentration is kept at a predetermined level by supplying carbon dioxide from the gas cylinder 14 as needed. Then, the suction is stopped once after a predetermined amount of the contaminated soil 17 and the contaminated water 4 are suctioned. They are stirred for a predetermined time in the separation vessel 10. When the cesium ions are sufficiently dissolved in the carbonated water 49, the on-off valve 18 is closed and the suction hose 11 is wound back around the reel 16, completing the suction work of the contaminated soil 17 and the contaminated water 4. Then, the on-off valve 51 is opened, and the switching lever 63 is switched from the suction position to the discharge position. With the above operation, the air is taken from the vent pipe 61 and then sent to the communication pipe 52 via the loop conduits 58, 59. The air is discharged from the top of the separation vessel 10 and whereby the separation vessel 10 is pressurized. Accordingly, the contaminated soil 17 and the contaminated water 4 are sent to the solid-liquid conduit 47 from the on-off valve 51 together with cesium ions separated from the tritiated water. These solid and liquid components are introduced into the solid-liquid separation filter 53. In the solid-liquid separation filter 53, the centrifuge is started before or after switching the switching lever 63. The solid-liquid components are introduced into the rotating tube, and the contaminated soil 17 is moved to the outer side of the rotating tube, while the carbonated water 49 which does not include no soil 17 is moved to the inner side of the rotating tube by their specific gravities, whereby solid-liquid separation is performed. At that time, the carbonated water 49 adhered to the soil 17 is separated from the soil 17 together with radioactive cesium ions by centrifugal effect. Accordingly, about the total amount of radioactive cesium ions and tritiated water are sent to the separation-liquid duct 56 together with the carbonated water 49 as the separated liquid 37. The soil 17a which does not include radioactive cesium ions is allowed to sink and accumulate in the lower part of the separation filter 49. This state is shown in FIG. 7. The separated liquid 37 is moved to the adsorption filter 42 provided at the top of the electrolysis tank 13 through the separation-liquid duct 56. The metal ions such as the radioactive materials and heavy metals are adsorbed by the filter 42 and the separated liquid 37 is flowed down to the airtight container 30 to soak the cathode 31 and deposition members 36 inside. After the separated liquid 37 is introduced into the electrolysis tank 13, the suction pump 15 is stopped and then the on-off valve 57 is closed. After checking the acid concentration of the separated liquid 37 with the pH sensor 43, carbon dioxide is supplied to the electrolysis tank 13 as needed to adjust the acid concentration of the carbonated water 49. At the same time, a heater (not shown) attached to the outside of the electrolysis tank 13 is turned on to promote electrophoresis of radioactive cesium ions. This state is shown in FIG. 7. Then, the switch 34 is turned on to pass a current between the electrodes 30, 31 to perform electrolysis of the separated liquid 37. Oxygen is generated in the airtight container 30, which is an anode, and hydrogen is generated at the cathode 31. Oxygen is generated from the side surfaces in the airtight container 30, and the air bubbles move up along the side surfaces and accumulate in the upper space in the collector 35. The insulating coating 30a is provided at the bottom of the airtight container 30 so that oxygen is not generated from the insulating coating 30a, thereby collecting the hydrogen safely. As described above, oxygen accumulated in the upper space in the collector 35 reaches a predetermined pressure, the safety valve 72 automatically opens and releases the oxygen to the outside through the discharge pipe 73. The air bubbles of the generated hydrogen float in the separated liquid 37 along the cathode 31 and the deposition member 36, and moved to the liquid level and collected in the collector 35. As the hydrogen is collected, the air in the collector 35 is pushed out. After the hydrogen is concentrated to the predetermined concentration, the suction pump 40 is started and the hydrogen in the collector 35 is suctioned and introduced to the trap 38. The hydrogen is then injected and charged into the gas cylinder 41 equipped with the end of the trap 38. This state is shown in FIG. 8. When the separated liquid 37 is electrolyzed, the radioactive materials and the metal ions such as the heavy metals dissolved in the carbonated water 49 travel toward the side of the cathode 31 and the peripheral deposition components 36 by electrophoresis and then deposited and attached thereon. In this case, the deposition members 36 are provided densely to surround the cathode 31. Thus, the metal ions are deposited precisely and reliably on the deposition members 36 and the cathode 31, and foreign matters in the separated liquid 37 are adsorbed and cleaned. This state is shown in FIG. 8. As described above, by charging hydrogen into the gas cylinder 41 and depositing the radioactive materials and the metal ions such as the heavy metals on the cathode 31 and the deposition members 36 for a predetermined period of time, a predetermined amount of the radioactive materials accumulate in the electrolysis tank 13. As a result, the electrolysis tank 13 needs to be disposed of and replaced. Then, the suction pump 15 is actuated, and the switching lever 63 is switched to the suction position to close the on-off valves 18, 46. The on-off valves 51, 57 are opened to suck muddy water containing the carbonated water 49 in the separated liquid 10 via the communication pipe 62, thereby creating a negative pressure in the separation vessel 10. The clean separated liquid 37 in the separation vessel 13 is suctioned with the separation-liquid duct 56 and then refluxed to the separation vessel 10 via the solid-liquid separation filter 53 and the solid-liquid conduit 52. Accordingly, the clean separated liquid 37 can be effectively used. This state is shown in FIG. 10. The adsorption filter 42 placed in the electrolysis tank 13 adsorbs the metal ions such as the radioactive materials and heavy metals, whereby a predetermined amount of the radioactive materials accumulate. Then, the used adsorption filter 42 is stored together with the electrolysis tank 13 in the storage facilities 67. The gas cylinder 41 charged with the predetermined amount of hydrogen is removed from the trap 38 and then stored in the storage facilities 69. A predetermined amount of the soil 17a which has been separated may accumulate on the solid-liquid separation filter 53 by the solid-liquid separation and this may become a problem for a next use. In such a case, the soil 17a is collected from the solid-liquid separation filter 53. In this case, the discharge valve 54 is opened and the soil 17a is dropped from the discharge pipe 54. The collected soil 17a is then dried. A specified soil activator 65 is added to the dried soil 17a to improve or modify the soil 17. The soil activator 65 is selectable from organic fertilizers such as compost, mycorrhizal fungi, or various kinds of chemical fertilizers including nitrogen, phosphorus, and potassium. Such soil activator 65 is added to and mixed with the soil 17a, and the soil 17a is returned to the original field 3 in which the soil has been collected. This state is shown in FIGS. 4 (g) and (h). In some embodiments, the powder fire extinguishant used as a soil activator 65 may be one for which the expiry date has passed. The powder fire extinguishant contains monobasic ammonium phosphate or ammonium sulfate. The hydrophilic fertilizer is made by using the powder fire extinguishant, whereby the fire extinguishant can be used effectively despite its expiry date. Accordingly, the soil 17a which has been improved or modified is returned to the field 3 from which the contaminated soil 17 is collected. The soil is improved and becomes more fertile than the original state. Thus, farming can be restarted more promptly in this method than the case where the contaminated soil is simply decontaminated and returned to the field 3. Meanwhile, clean separated liquid 37 is refluxed into the separation vessel 10, fresh water 24 in the water supply tank 12 is supplied to the separation vessel 10 as needed, and the dioxide is supplied from the gas cylinder 14 to prepare the carbonated water 49. Then the suction hose 11 is unreeled and the suction pump 15 is actuated to restart suction and collection of the contaminated soil 17, the contaminated water 4, and tritiated water mixed therein. Then, the contaminated soil 17, the contaminated water 4, and tritiated water are introduced into the separation vessel 10 in the same manner as described above. The metal ions containing the radioactive materials are dissolved into the carbonated water 49 and then the solid-liquid separation liquid is introduced to the solid-liquid separation filter 53 through the solid-liquid separation duct 56. The eluting solvent containing the soil 17a, the radioactive materials, and metal ions is separated into solid and liquid, and the separated soil 17a accumulates in the solid-liquid separation filter 53. The separated liquid 37 containing the radioactive materials, the metal ions, and the carbonated water 49 is introduced to the adsorption filter 42 to adsorb radioactive cesium, heavy metals in the separated liquid 37. As described above, in some embodiments, the remaining liquid of the separated liquid 37, the radioactive materials and metal ions deposited on the cathode 31 and the deposition member 36, the adsorption filter 42, and the safety valve 72 in the electrolysis tank 13 are stored collectively. Accordingly, they can be stored efficiently, compactly, and safely compared with a case in which the components are disassembled and stored individually. Additionally, an accident of radiation exposure can be prevented. After the used electrolysis tank 13 is removed, a new electrolysis tank 13 is insulated and installed at the same position on the decontamination vehicle 9. One end of the separation-liquid duct 56 is connected to the solid-liquid separation filter 53 and the other end is connected the on-off valve 57. One end of the trap 38 is placed in the airtight container 30, and the other end is connected to the suction pump 40. A discharge pipe of the suction pump 40 is inserted in a new gas cylinder 41. Additionally, one end of the lead 32 is connected to the top end of the cathode 31, and the other end of which is connected to the airtight container 30. Then the pH sensor 43 is replaced by attaching it to the airtight container 30. As described above, the electrolysis tank 13 is efficiently replaced at low cost since the component which has not been contaminated with radioactive material is used. The gas cylinder 41 charged with hydrogen is stored safely in the storage facilities 69. This state is shown in FIG. 13. In this case, lithium, a small amount of which is contained in hydrogen, has a half-life of 12.32 years, which is relatively short. Before the half-life period has passed, lithium needs to be tightly stored. After the half-life period has passed, the lithium turns into helium 3 (3He). Thus, helium 3 can be reused, or discharged into the air. As described above, in some embodiments, the decontamination vehicle 9 is driven to the object area to be decontaminated 1. The contaminated soil 17 and the contaminated water 4 in the object area 1 are collected and then decontaminated speedily with the facilities provided on the decontamination vehicle 9. The soil 17a which has been decontaminated is returned to the original field 3. A series of decontamination operations can be performed in the object area 1, and thus, the above decontamination work can be carried out efficiently and speedily, thereby promoting restart of farming and reducing the amount of the contaminated soil 17. Further, in the present embodiment, the invention can be applied not only to the contaminated soil 17 but to decontamination for the paddy field 5 and the wetlands where contaminated water 4 exists. Accordingly, the invention can be employed in wide areas and has practical effects. In the present embodiment, in addition to radioactive cesium, tritiated water mixed in the contaminated water 4 collected from the object area to be decontaminated 1 is separated from the soil 17. The separated liquid 37 used is electrolyzed in the electrolysis tank 13. The hydrogen generated by the electrolysis is collected and trapped, and then charged into the gas cylinder 41. Additionally, tritium, a small amount of which is contained in hydrogen, is filled in the gas cylinder 41 and stored safely in the storage facility 67. Accordingly, radiation exposure by tritium can be prevented. Accordingly, anxiety for developing cancer by DNA damage or gene damage due to internal exposure of tritium can be relieved. In the present embodiment, the used electrolysis tank 13, the adsorption filter 42, and the remaining liquid of the separated liquid 37 are sealed in the airtight container 30 so that they are stored safely. Accordingly, radiation exposure from the electrolysis tank 13 and the adsorption filter 42 is prevented, and the electrolysis tank 13 and the adsorption filter 42 are stored rationally and safely. As described above, the decontamination method and system for soil and the like of the present invention decontaminate soil of, for example, fields, and water contaminated with radioactive materials reliably and rapidly on site, aiming to perform decontamination with precision and improved efficiency. According to the present invention, the decontaminated soil is improved by adding soil activators, and then the improved soil is returned to the original field readily to promote the restart of farming. Further, according to the present invention, the radioactive materials adhering to or deposited on the soil are separated from the soil and then concentrated precisely. According to the invention, a reduction in the volume of the contaminated soil and a safe treatment of the radioactive materials are achieved. Further, according to the invention, decontamination of radioactive cesium and tritium is achieved, thereby relieving anxiety of internal exposure to the radiation of tritium. Further, the decontamination apparatus can be disposed of efficiently and safely.
	abstract	A system for the distributed diagnosis of a physical system includes serveral local diagnostic subsystems that generate local diagnoses based on observations of a component of the physical system. An interface is defined by which the local diagnostic subsystems communicate. Further, an algorithm for assembling a global diagnosis from the local diagnoses is fined.
	abstract	The disclosed CT scanner comprises at least one source of X-rays; a detector array comprising a plurality of detectors; and an X-ray filter mask arrangement disposed between the source of X-rays and detector array so as to modify the spectra of the X-rays transmitted from the source through the mask to at least some of the detectors so that the X-ray spectra detected by at least one set of detectors is different from the X-ray spectra detected by at least one other set of detectors.
	abstract	The disclosed invention proposes a reconfigurable radiation shield that, compared to art static shields, improves the protected volume/weight ratio. The reconfigurable shield is applicable in the medical field, in the aerospace industry, in mobile radiological laboratories and decontamination vehicles, as well as in other fields where intensity-fluctuating radiation and variable direction radiation represent a hazard.
044514271	description	DESCRIPTION OF THE PREFERRED EMBODIMENT As set forth in the Background of the Invention, in-core fuel management is a very important feature of nuclear power plant design. Therefore, much time and money is spent by nuclear reactor vendors to optimize the fuel management for each particular reactor through the use of detailed computer simulation prior to fuel fabrication. All data presented in the following description of the invention were generated in the course of a computer simulated verification that the inventive concept would indeed satisfy the above-recited objectives of the invention. The calculational models for implementing the invention are well-known in the art of nuclear reactor fuel management, and the following description used in conjunction therewith will enable one ordinarily skilled in this art to adapt the invention for use in any size PWR for any fuel cycle requirements ordinarily desired for large electric power generating stations. In one embodiment, the invention is implemented in a reactor core that has previously been loaded with fuel for one or more cycles according to some prior art scheme. Such an embodiment is illustrated in FIGS. 2(a) and 4, where a second cycle embodying the invention immediately follows the prior art OI first cycle shown in FIG. 3. The following table, used in conjunction with FIG. 3, summarizes the important fuel design properties of the OI first cycle, and will serve as a reproducible starting point for practicing the embodiment of the invention described hereinbelow. TABLE 1 ______________________________________ Fuel Design for First Cycle Prior Art Out-In Scheme Shown in FIG. 3 Enrich- Shim No. ment No. Loading Assembly Shims in No. As- (wt. % Fuel (wt. % B.sub.4 C in Type Assembly semblies U-235) Rods B.sub.4 C--Al.sub.2 O.sub.3) ______________________________________ A 0 81 1.83 19116 -- BL 16 36 2.49 7920 2.76 BH 16 52 2.49 11440 3.37 CL 16 24 2.95 5280 2.04 CH 16 8 2.95 1760 3.37 C 0 40 2.95 9440 -- ______________________________________ In FIG. 3, the numeral 16 in the upper left corner of each assembly identifies an assembly location. Location number 69 is at the core center, and the parts of the core not shown and the fuel contained therein are merely reflections along the major axes. A schematic of a typical fuel assembly 18 is shown in FIG. 5, where fuel rods 20, fixed burnable poison lattice shims 22, water holes 24, and guide tubes 26 (one shown) are represented. More details of the core and fuel assembly designs can be found in the Combustion Engineering Standard Safety Analysis Report (CESSAR) Docket No. STN-50-470 Section 4.3 (1975), which is incorporated by reference. In order to more clearly compare and distinguish the invention from the prior art, FIGS. 1(a) and 1(b) show how the prior art would be used to design a second cycle following the same 13,800 MWD/T first cycle shown in FIG. 3. In the OI prior art second cycle scheme shown in FIG. 1(a), the A assemblies are removed (except that the most reactive A assembly is moved to the core center), the B fuel 14 and C assemblies 12 relocated as shown, and unshimmed D assemblies 10 having an average enrichment of 3.50 wt % are inserted. The beginning of cycle 2 (BOC2) core average initial enrichment is 2.97 wt %, sufficient for a second cycle burnup of 10,000 MWD/T. In the IOI cycle 2 prior art scheme shown in FIG. 1(b), the D assemblies 10' are shimmed and have an average enrichment of about 2.96 wt %. The core average BOC2 enrichment is about 2.75 wt % for the same energy extraction as the OI scheme. The second cycle scheme embodying the present invention is shown in FIGS. 2(a) and 4. The inventive concept contained therein is derived from the discovery that the inner checkerboard in the IOI scheme of FIG. 1(b), which has one component of feed fuel 10 (L) and another component of B (L-2) fuel 14, can be significantly violated yet give an overall improvement in the gross power distribution and a decrease in the required shim worth, by an interchange of feed (L) fuel 10 and C (L-1) fuel 12 according to a general procedure to be described below. The resulting new in-core fuel management scheme can be characterized by reference to an imaginary boundary 28 between an inner region 30 containing about two-thirds of the assemblies and an outer region 32 as shown in FIGS. 2(a) and 2(b). The recommended outer boundary of the inner region consists of all assemblies intersected by a circle drawn about the core center, perpendicular to the vertical axis of the core and having a radius equal to three-quarters the distance from the core center to the closest point on the outer edge of the core periphery. In FIG. 2(b), the distance to the periphery is labeled P and the boundary circle radius is labeled R. The following table summarizes the feed fuel assembly properties represented in FIG. 4. The numeral 34 in FIG. 4 indicates the previous location of the A, B, and C assemblies. The numeral 36 in the lower right corner of the D assemblies indicates the type of shim loadings and distribution resulting from application of the method to be described below. TABLE 2 ______________________________________ Feed Fuel Design for Second Cycle Using the Invention As Shown in FIG. 4 Assembly No. Shims No. of Type per Assembly Assemblies Enrichment Shim Loading ______________________________________ D401 0 32 3.28 0 D402 8 16 3.01 1.59 D403 4 8 3.01 1.82 D404 8 8 3.01 1.87 D405 8 4 3.01 2.21 D406 8 8 3.01 1.99 D407 4 4 3.01 3.12 ______________________________________ The following is a detailed description of the method of implementing the invention. The intent is to satisfy certain reactivity and power relationships at the beginning of each cycle, which have been found to consistently produce, particularly at EOC, the advantages of the invention as described above. This method instructs one to arrange fuel at BOC by first determining what the limiting K infinite (hereinafter K) balance in the core can be at EOC and still satisfy the local fuel rod peaking limits, then working back to the feed assembly enrichment, shim strength, and placement that will, with burnup, come within the EOC K balance. The steps in the method are based more on the characteristics of the core and fuel assembly design than on the specific fuel management scheme used the prior cycle. Thus, one familiar with the basic core and fuel assembly characteristic of a particular reactor in which prior art fuel management techniques have been used, can with relatively little effort implement the present invention. First, the core geometry is divided into an inner region which contains approximately two-thirds of the assemblies, and an outer region containing the remainder of the assemblies. A recommended boundary between the regions is a circle about the core center having a radius equal to three-quarters the shortest distance from the core center to the core periphery. From previous, commonly available calculations, the ratio of the hottest fuel rod in the inner region 30 to the average rod in the inner region is determined. This ratio, Pi/Pi, is preferably obtained from existing fuel management schemes which use the present invention or the IOI technique, but OI power distributions can be used if the calculated ratio is augmented by the ratio of power of an EOC feed assembly to the power of an adjacent EOC twice-burned assembly. This augmentation factor can be determined from a checkerboard calculation having typical end of cycle fuel characteristics. The next step is to determine the relationship at EOC of the difference in K between the outer region 32 and the inner region 30 (.DELTA.K.sub.o-i), and the resulting ratio of the average power in the inner region to the core average power, Pi/P. FIG. 6 shows this relationship for the 241 assembly and the 217 assembly cores shown in various other figures, where the basic fuel assembly design shown in FIG. 5 is employed. This relationship is determined from surveying several end of cycle power distributions from any fuel management scheme wherein the absorption of all shim poison material is cancelled from the calculation so as to represent zero shim residual. The designer then chooses the design target axially integrated radial peak fuel rod to core average rod power ratio, commonly known as F.sub.r, a value usually imposed on the designer as a consequence of safety considerations. By dividing F.sub.r by the ratio Pi/Pi, the maximum permitted value of Pi/P consistent with the design target F.sub.r is obtained. In the present example, F.sub.r is 1.41 and Pi/P is 1.28. The required division indicates a permitted inner region power ratio Pi/P of about 1.10. Referring again to FIG. 6, it can be seen that the end of cycle difference .DELTA.K.sub.o-i (EOC) required to produce a Pi/P equal to 1.10 is 9.2%. In order to obtain the same K difference at beginning of cycle .DELTA.K.sub.o-i (BOC) to assure an F.sub.r less than 1.41 based on the difference in K determined immediately above .DELTA.K.sub.o-i (EOC), a correction must be made for the difference in regionwise exposure between end of cycle and beginning of cycle. The first step is the determination of the difference in accumulated exposure between the inner and outer regions over the cycle. This difference is just (Pi/P-Po/P)* CYCLE LENGTH. In the present example where Pi/P is 1.1, Po/P for the outer one-third core is 0.8, and for a cycle length of 10,000 MWD/MTU, the inner region accumulates an additional 3,000 MWD/MTU relative to the outer region. This difference between inner and outer region exposure is converted into a reactivity difference according to well-known derivatives of the change in core K with exposure. In the present example, the adjusted reactivity difference (unshimmed) at BOC2 is about 6.5%. This BOC2 .DELTA.K must be further adjusted to account for the shim residual poison carried over from the EOC1 batch B and a few C fuel assemblies. This EOC1 shim residual poison is depleted during the course of cycle 2 and does not contribute to the difference in regionwise K at the EOC 2. The adjusted reactivity difference at BOC2, allowing for the shim residual carried over from cycle 1, is about 7.5% .DELTA.K. As will be described below, the difference between this BOC2 value and the EOC2 regionwise reactivity difference of 9.2% .DELTA.K is accounted for in the design through the placement of shims in the fresh assemblies. It is the latter reactivity difference, of 9.2% .DELTA.K, that the designer strives for an order not to exceed a peak fuel rod power F.sub.r of 1.41. Experience shows that in a scheme arranged with the present method, the absolute value of the peak and the regionwise power density Pi/P remain fairly constant throughout the burnup cycle. The next step is to make a rough estimate of the required fresh feed enrichment in the D batch, which can be obtained by taking the core average initial enrichment required to produce the desired second cycle length using the IOI scheme and adding about 0.15 wt %, or using the OI scheme and subtracting about 0.4 wt %. In the present example, the D feed enrichment is about 3.12 wt %, and the beginning BOC2 core average enrichment is about 2.85 wt %. At this point, the following target characteristics have been estimated for BOC2: the reactivity difference between the outer and inner regions (9.2%), the amount of this reactivity difference that should be distributed as shims in the fresh assemblies in the inner region (1.7%), and the average enrichment of the fresh fuel (3.12 wt %). It remains to choose the specific shim loadings (boron content) for the fresh assemblies, and to arrange all the assemblies in the reactor core. This can be facilitated by performing a few preliminary trial and error hand calculations of .DELTA.K.sub.o-i (BOC) based on known values of K for each fuel assembly in the core at BOC2. The assemblywise K's can be obtained by performing a single core reactivity calculation at the estimated BOC2 soluble boron concentration with no xenon and peak samarium in the burned (L-1, L-2, . . . L-N) assemblies. Fresh assemblies having a variety of shim loadings are included in this calculation, so that a relation between K and shim loading is determined. The adequacy of specific shim loadings and fuel assembly arrangements can be estimated through trial and error according to the following plan. The inner region of the core is filled with a quarter core symmetric checkerboard having one component of L and a second component of L-2 assemblies. L-1 assemblies are placed toward the core periphery. An arithmetic reactivity difference is calculated between the outer and inner regions of the core, which will generally be smaller than the target .DELTA.K.sub.o-i (BOC). Shims are located in L assemblies such that the inner region contains about 2.7% more shim worth than the outer region (1.7% in fresh assemblies and 1.0% carryover from first cycle). The inner region reactivity must be further reduced, and this is accomplished through the key step of interchanging L-1 assemblies from the outer region with L assemblies from the inner region. It will be generally found advantageous to place L assemblies in several peripheral locations. Thus in the embodiment illustrated in FIGS. 2(a) and 4, for example, the core periphery consists only of L and L-1 assemblies, and in particular no more than half the core periphery contains L assemblies. The L-1 and L assemblies are interchanged, and the shim loadings and placement are manipulated, until the hand calculation indicates the desired .DELTA.K.sub.o-i (BOC) (9.2%) and the desired L assembly shim worth in the inner region (1.7%) have been achieved. At this point, customary computer calculations can be employed to fine-tune the power distribution and to verify the estimated enrichment. FIG. 4 and Table 2 include information showing the resulting change in location of the L-1 and L-2 (and a single L-3) assemblies from EOC1 to BOC2. Also shown are the number of shims in each L assembly and the shim loading in wt % of B.sub.4 C (containing natural boron) in B.sub.4 C-AL.sub.2 O.sub.3 shim material. The invention is not limited to use with B.sub.4 C shim material, however, and can be practiced, for example, with lattice shims composed of an admixture of gadolinium and fuel material (UO.sub.2), or with removable shims whether or not located in the guide tube. It is well within the skill of an ordinary nuclear fuel management engineer to substitute other shim material, or other fuel lattices, without departing from the scope of the invention. It is to be understood that once the target BOC arithmetic .DELTA.K.sub.o-i difference is achieved, a computer calculation of the power distribution during the cycle is to be made. It is expected that several iterations in which minor adjustments of shim loadings, fuel enrichment, or fuel assembly placement are made may be needed before satisfactory power distributions and EOC reactivity are obtained. After practicing the present invention a few times, however, one having ordinary skill will need only about two or three such iterations. Referring again to FIG. 2(a) the differences in the arrangement of fuel assemblies with the present embodiment of the invention can be identified relative to the arrangements of the prior art OI scheme shown in FIG. 1(a) and the IOI scheme shown in FIG. 1(b). With respect to the boundary between the inner and outer region indicated by a heavy line 28, the present invention consists of a checkerboard pattern in the inner region having one component consisting of L (10, 10') and L-1 (12) assemblies and a second component consisting of L-2 assemblies (14). The core geometry of FIG. 2(a) is shown in FIG. 2(b) where first component 40 and second component 42 lines of the inner checkerboard and third component 44 and fourth component 46 lines of the outer checkerboard (to be later described) are indicated. The prior art does not show a checkerboard wherein the first component 40 consists mostly of L and L-1 assemblies. In the embodiment shown, the second component 42 of the inner checkerboard consists entirely of L-2 fuel and, when the center assembly is included, L-3 fuel. It is also seen that the outer region 32 consists of a checkerboard of L assemblies on the third component 44 alternating with a fourth component 46 of L-1 and L-2 assemblies. The OI scheme of FIG. 1(a) intentionally avoids checkerboarding L fuel in the outer region. There is no discernable checkerboard pattern in the outer region of the IOI scheme shown in FIG. 1(b), since adjacent components of L-1 fuel near the periphery have no L fuel. With respect to the OI scheme of FIG. 1(a), none of the L assemblies is shimmed, whereas in the present invention at least some of the L assemblies 10' are shimmed. Furthermore, in the present invention less than two-thirds of the L assemblies are in the outer region, whereas in the OI scheme almost all L assemblies are in the outer region. With respect to the IOI scheme shown in FIG. 1(b), no L assemblies are on the core periphery whereas in the present invention there are several L assemblies on the periphery. Furthermore, every L assembly 10' is shimmed in the IOI scheme, whereas in the present invention the outer region includes at least some unshimmed assemblies 10. The above comparison of the present invention with the prior art is based on the preferred embodiment of the invention. As will be described below, different fuel management objectives may require different relative fractions of L, L-1, L-2, . . . L-N assemblies in the core, and the checkerboard components may therefore not be as perfectly filled as in the present embodiment. Nevertheless, the essential characteristic of the present invention, the first component of the inner checkerboard consisting mostly of L and L-1 fuel, is found in all embodiments of the invention. FIG. 7 shows the invention practiced in the first cycle of a core having 217 assembly locations. The core contains unshimmed A fuel 14, shimmed (BS) 12' and unshimmed B fuel 12 and shimmed (CS) 10' and unshimmed C fuel 10. In this embodiment, the first component 40 of the inner region consists of L (10, 10') and L-1 (12, 12') assemblies, and the second component 42 consists of L-2 (14) assemblies. In the outer region the third component 44 is chosen from L, L-1 assemblies and the fourth component 46 is chosen for L, L-1, and L-2 assemblies. Although a few minor modifications are required to the outline of steps discussed previously for implementing the inventive scheme, an ordinarily skilled nuclear reactor fuel management engineer can easily adapt the above procedures for use in designing the first cycle. For example, it is well known that in the first cycle most or all of the B (L-1) as well as the C (L) assemblies require substantial shim loadings. FIG. 8 shows a later cycle scheme in the 217 assembly core in which the first component of the inner checkerboard contains four L-2 assemblies in each quadrant (assembly locations 16, 23, 43 and 51). This deviation from a perfect L and L-1 first component is sometimes the best way to accommodate peculiarities of the core power distribution in which certain assembly locations exhibit high power peaks. It is believed that a minimum of two-thirds of the first component locations must contain L and L-1 assemblies, and that at least one-third of all L assemblies be in the inner region, in order not to depart from the inventive concept. It is noted that it may not be necessary to use shims in every L assembly of the first component, especially if several different enrichments are used in each batch. This would permit concentrating the desired shim worth in only a few L assemblies in the inner region. Although such an arrangement falls within the scope of the invention, it is believed that the power distribution cannot be controlled if more than one-third of the L assemblies in the inner region are unshimmed. Referring now to FIGS. 9, and 10, there is shown an application of the present invention in the 241 assembly core designed for fractional batch fuel cycles. In fractional batch management, the distinction between a lot of fuel and a batch of fuel becomes important. In the normal three batch fuel management, a batch and a lot are synonymous because all the assemblies in a batch remain in the core for the same number of cycles and are removed together. In the fractional batch scheme shown in FIGS. 9, and 10, some assemblies of a batch are removed while others remain in the core for the next cycle. For example, in the third cycle fractional batch scheme shown in FIG. 10, the L or feed lot, 10, 10' contains 92 assemblies, the L-1 lot 12 contains 92 assemblies, and the L-2 lot 14 contains only 57 assemblies. This means that before the L-1 assemblies are shuffled for the next cycle, 35 are permanently removed from the reactor, leaving only 57 as L-2 assemblies. In the second cycle shown in FIG. 9, the first component 40 of the inner checkerboard consists of L and L-1 assemblies, and the second component 42 consists of L-2 assemblies. In the outer region, the third component 44 consists of L and L-1 assemblies and the fourth component 46 comprises assemblies chosen from lots L-1 and L-2. In the third cycle embodiment shown in FIG. 10, the first component 40 of the inner region checkerboard consists of L and L-1 assemblies, and the second component 42 consists of L-1 and L-2 assemblies. In the outer region checkerboard, the third component 44 consists of L and L-1 assemblies and the fourth component 46 consists of L-1 assemblies. It can be appreciated that, as fuel management schemes become more tailored to the individual needs of particular utilities, the use of fractional batch fuel management will be more common. Nevertheless, the present invention finds application in such use and the procedures outlined above for implementing the inventive scheme can easily be adapted for use with the more complex schemes.
053234316	summary	BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates generally to nuclear reactor vessels and, more particularly, to a device for securing a reactor vessel washer to a reactor vessel stud, allowing the stud and washer to be maintained as a mated pair during removal and replacement of the stud. 2. Description of the Prior Art In a commercial nuclear reactor vessel, heat, from which steam and ultimately electricity are generated, is produced by fission of a fissile material, such as uranium disposed inside the reactor vessel. The reactor vessel includes a body having a cylindrical wall with two ends. The wall includes an outwardly extending flange at one end, and a semicircular domed bottom is integrally attached to the opposite end of the wall. A reactor dome is attached to the body and is positioned atop the body. The dome has a semicircular shape with an annular shaped lip portion as is well known in the art. An outwardly extending flange is disposed around the circumference of the dome at the lip portion. The body flange is positioned abutting the dome flange. A plurality of holes through the dome flange are in registry with a plurality of holes inserted partially through the body flange. Each hole in the dome mates with a corresponding hole in the body flange forming a matched pair of holes. A plurality of washers, in registry with a flange hole, are positioned over the matched holes. A plurality of studs each having two ends and a threaded shaft are used to attach the dome to the body. The plurality of studs, each with a nut attached at one end, are disposed respectively through the matched holes with the stud and washer forming a mated pair. The studs are threaded into both flanges such that one end passes through both flanges, and the other end, which accepts the nut, extends upwardly from the abutting flanges. The stud is disposed in a matched pair of holes such that the nut abuts the top of the washer. The nut and washer together function to rigidly attach the dome to the body. Due to maintenance procedures, regulations, and the need to replenish expended fuel, the interior of the reactor vessel should be periodically inspected. To inspect the interior, the dome may be repeatedly removed. However, once the dome is initially installed, during reinstallation of the dome the stud and washer should be maintained as a mated pair. This is because the stud is stressed when it is installed, and the nut and washer, thereafter, conform to the stressed stud. Therefore, each stud, nut and washer is slightly altered and installing them on a different stud could cause them to fit improperly. One means of removing the reactor vessel head to facilitate inspection of the interior of the vessel is disclosed in U.S. Pat. No. 4,873,760. This patent discloses an apparatus which includes a gripper and is positioned over the dome. The gripper unthreads a nut associated with a particular stud, and then an attracting device magnetically attracts the washer such that the washer is forced into the gripper and abuts the nut. The gripper places the nut and washer on a pin and repeats the process for each of the nuts to be removed. The studs are then unthreaded, and the dome is removed. Another known method includes an apparatus, generally referred to in the art as an Automated Stud Handling and Transportation System, which is positioned on the vessel head flange. A stud remover of this apparatus completely removes the stud with the nut secured on one end of the stud. The washer is manually removed from the stud and affixed to a hook on the apparatus. Next, a movable arm for holding the stud attaches to the stud. The stud remover then moves so that it is positioned over an adjacent stud. The above process is repeated until all studs are removed. Once all the studs are removed, the dome is removed by a crane. To install the studs, the process is reversed. A major drawback to each of the above methods is that the stud and washer are removed separately. Thus, repeatedly removing and installing the studs makes the identification of which washer should be associated with a particular stud extremely difficult. As stated above, the same nut and washer should be maintained as a mated pair each time the stud is removed and then replaced. Prudent practice dictates that if a particular stud and its associated washer and nut can't be identified, a new nut and washer should be installed. This obviously results in increased costs each time the vessel head is removed and thereafter re-installed. Another drawback to each of the above methods is that when installing the studs a worker must carefully position each washer concentrically over the matched holes. This is a time consuming operation, and, therefore, a worker will have increased exposure to radiation in the reactor. Still further, if the washer is improperly aligned with the matched holes, the washer may be damaged as the stud is inserted. Consequently, a need exists for a simple device for removably securing a reactor vessel washer to a vessel stud allowing the washer and stud to be maintained as a mated pair during removal and replacement of the stud. SUMMARY OF THE INVENTION The present invention provides an improvement designed to satisfy the aforementioned needs. Particularly, the present invention is directed to a device for removably securing a reactor vessel washer to a reactor vessel stud and is operable for use with a reactor vessel including a dome having a first flange positioned abutting a second flange of a reactor body, a plurality of reactor studs are disposed in both the first and second flanges; and a nut and washer, both having an inner peripheral surface, used for tightening the studs and for attaching the dome to the body. The device comprising a retainer removably disposed on the stud and operable to mate the washer to the stud to maintain the positional relationship of the stud and washer as the stud is removed from the dome and body.
	claims	1. A detector assembly including:a semiconductor detector having a first surface and a second surface opposed to each other, the first surface including pixels comprising pixelated anodes, and the second surface comprising a cathode electrode;a collimator including openings, each opening associated with a single corresponding pixel of the semiconductor detector; anda processing unit configured to identify detected events within virtual sub-pixels distributed along a length and width of the semiconductor detector, wherein each pixel comprises a plurality of corresponding virtual sub-pixels, wherein absorbed photons are counted as events in a corresponding virtual sub-pixel, wherein absorbed photons are counted as events within a thickness of the semiconductor detector at a distance corresponding to an energy window width used to identify the events as photon impacts. 2. The detector assembly of claim 1, wherein an absorption location for each absorbed photon within the thickness of the semiconductor detector is defined within a range such that ΔL/D=ΔE/E, where ΔL is a distance from the cathode, D is the detector thickness, ΔE is an energy window width, and E is a photopeak energy of an absorbed photon. 3. The detector assembly of claim 1, wherein the processing unit is configured to determine an absorption location for a given absorbed photon based on non-collected signals received from pixelated anodes adjacent to a pixelated anode absorbing the given absorbed photon. 4. The detector assembly of claim 1, wherein the collimator is a pinhole collimator. 5. The detector assembly of claim 1, wherein the collimator is a parallel-hole collimator. 6. The detector assembly of claim 1, wherein the events for each pixel are counted at a single shared location. 7. The detector assembly of claim 1, wherein the events for each pixel are counted within a single shared range. 8. The detector assembly of claim 1, wherein each pixel comprises virtual sub-pixels along the length and width, but not along the thickness of the semiconductor detector.
	claims	1. A radiation detector, comprising:an array of 2D collimators configured to collimate in at least two collimation directions, each of the 2D collimators of the array including 2D collimator modules arranged in series, adjacent ones of the 2D collimator modules being glued together via a layer of adhesive, to establish a fixed mechanical connection to facing module sides of the 2D collimator modules, relatively outer ones of the 2D collimator modules including at least one retaining element on at least one remaining side, the retaining element including a screw mechanism for mounting each of the 2D collimators of the array opposite a detector mechanism, whereineach of the 2D collimators of the array is replaceable independently of the remaining ones of the 2D collimators of the array. 2. The 2D collimator as claimed in claim 1, wherein the facing module sides are implemented such that an absorber surface of an absorber element of one of the 2D collimator modules, running parallel to the module side, is glued to edges of absorber elements of other adjacent ones of the 2D collimator modules. 3. The 2D collimator as claimed in claim 1, wherein the facing module sides are implemented such that absorber surfaces, running parallel to the module sides, of an absorber element of the adjacent 2D collimator modules are glued together. 4. The 2D collimator as claimed in claim 1, wherein, for mutual alignment of the adjacent 2D collimator modules, there is provided on one facing module side, at least one projection to engage in at least one recess in the corresponding other module side. 5. The 2D collimator as claimed in claim 1, wherein the at least one retaining element includes,at least one fastening device to fasten the respective 2D collimator to the detector mechanism; andat least one adjustment device to position the 2D collimator in the collimation direction with respect to the detector mechanism. 6. The 2D collimator as claimed in claim 5, wherein the at least one adjustment device for positioning the 2D collimator with respect to the detector mechanism in a radiation incidence direction includes a bearing surface which, when the 2D collimator is incorporated in the detector mechanism in the radiation incidence direction, comes to rest against a support surface of the detector mechanism. 7. The 2D collimator as claimed in claim 1, wherein at least the outer 2D collimator modules are manufactured in one piece with the at least one retaining element. 8. The 2D collimator as claimed in claim 7, wherein the 2D collimator modules are produced using selective laser sintering. 9. A method for manufacturing an array of 2D collimators collimating in at least two collimation directions with 2D collimator modules disposed in at least one collimation direction of each of the 2D collimators, the method comprising:preparing a plurality of 2D collimator modules;applying a layer of adhesive to at least one module side of adjacent ones of the 2D collimator modules;forming each of the 2D collimators of the array from a given number of the plurality of 2D modules, each of the 2D collimators of the array being replaceable independently of the remaining ones of the 2D collimators of the array;mounting each of the 2D collimators opposite a detector mechanism via at least one retaining element including a screw mechanism; andplacing the 2D collimators in a precision tool at a position provided for respective 2D collimator modules. 10. The method as claimed in claim 9, further comprising:gluing the at least one retaining element to at least one free side of relatively outer ones of the 2D collimator modules. 11. The method as claimed in claim 9, wherein the preparing includes producing the 2D collimator modules using selective laser sintering. 12. A radiation detector, comprising:an array of 2D collimators, the 2D collimators of the array comprising 2D collimator modules arranged in series, adjacent ones of the 2D collimator modules being glued together to establish a fixed mechanical connection to facing module sides of the 2D collimator modules, relatively outer ones of the 2D collimator modules including at least one retaining element on at least one remaining side, the retaining element including a screw mechanism for mounting each of the 2D collimators of the array opposite a detector mechanism, whereineach of the 2D collimators of the array is replaceable independently of the remaining ones of the 2D collimators of the array, andeach of outer ones of the 2D collimators being formed as one piece and including one 2D collimator module.
051035049	description	The T-shirt or shirt according to FIG. 1 is provided for persons wearing a heart pacemaker and is designed to shield the heart pacemaker against electromagnetic radiation, especially in the microwave range. For this purpose, the T-shirt is made of the fabric described in detail above which is woven from a mixed yarn of cotton and steel fibers intertwined with each other. The T-shirt completely covers the upper part of the body of the wearer, and clearly extends with its lower rim 6 beyond the hip area of the wearer. The neckline 2 of the T-shirt equipped with a folded collar 1 can be closed by means of a zip fastener 3 of plastic material, the zip fastener being underlayed with an interior flap 4 completely covering the neckline 2 even when the zip fastener 3 is completely open. The interior flap 4 is made of the same textile fabric material as the other parts of the T-shirt, whose sleeves 5 are elbow-length so that no detrimental radiation can intrude laterally from the sleeves 5, either. These sleeves 5 should be at least 20 centimeters long. The overall according to FIG. 2 may, for example, be used as working clothing for radar personnel and is also made of the textile fabric parts according to the invention. The overall can be closed up to the upper rim of the stand-up collar 7 by means of the zip fastener 3, is also made of plastic material, and is underlayed over its entire length with an interior border band (not shown) which is at least 7 centimeters broad and made of the textile fabric material. The pockets 10 are put on the textile fabric material and sewed to it. These pockets may be made of a different textile cloth. The two-piece protective suit according to FIGS. 3 to 9 is provided as protective clothing for hospital personnel exposed to microwave radiation arising from electromedical apparatus. The jacket, as well as the trousers of this protective suit, are made of an outer cloth such as light cotton fabric material developing no protective effect, and of a lining made of the textile fabric material according to the invention. This lining extends over the entire clothing. The front part of the jacket has a widened outer portion 14 extending up to one shoulder and an inner portion 13 extending with its interior border 17 to the middle of the chest. The jacket is closed by means of Velcro-type fastener strips 18 when laying together the outer portion 14 on the inner portion 13. The jacket 15 is provided with a stand-up collar 7 and elbow-length sleeves 5. The pair of trousers 16 can be closed in the front by means of a zip fastener 3 which is underlayed with a particularly wide interior border band 21 of a width of at least 5 centimeters. The waist measurement of the trousers is adjustable by means of Velcro-type fastener strips 18. In addition, the trousers have put-on pockets 10. The respective clothing according to FIGS. 1 to 9 are tailored of fabric pieces which are sewed together along joint seams 19. As can be seen from FIG. 10, the joint seams 19 are turned up into each other and sewed together by at least two seams 20.
063103559	claims	1. A shield for attenuating the flux of electromagnetic radiation from an article, the shield comprising: a flexible matrix comprising a foam including a radiation attenuating material, the matrix including at least one space within the matrix; whereby the at least one space reduces the weight of the shield without appreciably reducing the attenuating characteristics of the shield. providing a flexible matrix including a radiation attenuating material, and providing at least one layer including a space within the matrix; whereby the space reduces the weight of the covering without appreciably reducing the attenuating characteristics of the covering. 2. The shield of claim 1, wherein the shield has a transmission attenuation factor of at least 50% primary 100-kVP x-ray beam, a durometer of less than about 100 shore "00" and a coefficient of sliding friction relative to said article of at least 0.15. 3. The shield of claim 1, wherein the matrix is an expandable foam. 4. The shield of claim 1, wherein the matrix is an insulating material. 5. The shield of claim 1, wherein the matrix is a sponge. 6. The shield of claim 1, wherein a gas is provided in the at least one space of the matrix. 7. The shield of claim 1, wherein the attenuating material is barium sulfate. 8. The shield of claim 3, wherein the matrix includes silicone or urethane. 9. A method of making a covering for attenuating the flux of electromagnetic radiation, the method comprising: 10. The method of claim 9, further comprising providing the matrix as a foam matrix. 11. The method of claim 10, wherein providing the matrix as a foam matrix further includes injecting the foam matrix into the covering. 12. The method of claim 11, further comprising curing the foam matrix. 13. The method of claim 11, further comprising expanding the foam matrix. 14. The method of claim 9, further comprising providing the matrix as a gel. 15. The method of claim 9, wherein the layer comprises a groove. 16. The method of claim 9, wherein the layer comprises a slot. 17. The method of claim 9, wherein the layer is continuous. 18. The method of claim 17, wherein the layer is non-random. 19. The method of claim 11, wherein the layer is disposed between a first foam layer and a second foam layer. 20. The method of claim 19, further comprising injecting the foam. 21. The method of claim 20, wherein the radiation attenuating material includes bismuth. 22. A shield for attenuating the flux of electromagnetic radiation, the shield comprised of a flexible matrix comprising a gel and including an insulating material and a radiation attenuating material, the matrix including at least one space within the matrix, whereby the at least one space reduces the weight of the shield without appreciably reducing the attenuating characteristics of the shield. 23. The shield of claim 22, wherein the gel comprises a visco-elastic material. 24. The shield of claim 23, wherein the gel comprises a hydrogel. 25. The shield of claim 24, wherein the at least one space comprises a bubble. 26. The shield of claim 24, wherein the at least one space is disposed between a first layer of the gel and a second layer of the gel. 27. The shield of claim 22, further comprising a filler in the at least one space. 28. The shield of claim 27, wherein the filler comprises a wax.
062018469	summary	This invention relates to an improved method of jacketing bodies in pressure-tight jackets. More specifically, the invention relates to an improved method of jacketing a member comprising fissionable material within a jacket of nonfissionable material to form an assembly adapted for use in a neutronic reactor. In a neutronic reactor, a fissionable material, such as uranium, is commonly encased or jacketed in a jacket of a material of high thermal conductivity. The jacket functions both to prevent corrosion of the fissionable material and to prevent the escape of radioactive substances, which are generated within the fissionable material during operation of the reactor, into portions of the reactor other than the fissionable material. For operation of a neutronic reactor at high temperatures, the body within the jacket is bonded to the inner surface of the jacket. This is necessary in order to secure adequate heat transfer to an exterior coolant from the fissionable material, within which great quantities of heat are produced in the operation of the reactor. The fissionable members commonly employed in neutronic reactors are cylindrical in shape, usually of a diameter of the order of 1 inch and are of a length rendering them suitable for easy handling. Such fissionable elements are commonly called "slugs". It has been found that with the methods heretofore used for making the unitary jacketed slug structure described above, there have frequently occurred faults in the bonding between the fissionable member and the jacket. At the points of occurrence of such faults, there have developed "hot spots", points of high temperature, which may cause the production of blistering, rupturing of the jacket, and other deleterious effects upon the jacketed slugs during high temperature operation of the neutronic reactor. The development of such defects during operation is extremely prejudicial, if not fatal, to the continued proper operation of the reactor. Because of the factors discussed above, the quality and uniformity of the bond required between the fissionable member and the jacket must be perfect to a degree heretofore never required. The present invention is not directly concerned with the physical construction or the materials of the fissionable member, the jacket, or the bonding agent. It is the object of this invention to provide a method of assembling and bonding a jacketed slug of fissionable material wherein no air bubbles or other faults exist in the bond between the fissionable material and the jacket.
	abstract	The present invention relates to a passive filtration system for a fuel handling area having a spent fuel pool in a nuclear reactor. The passive filtration system reduces a discharge into the atmosphere of particulates, such as radioactive particulates, generated in a spent fuel pool boiling event. The passive filtration system includes a discharge path, a vent mechanism positioned between the fuel handling area and the discharge path. The vent mechanism is structured to release a steam and air mixture from the fuel handling area to the discharge path. The steam and air mixture includes the particulates. The passive filtration system further includes an air filtration unit located in the discharge path and this unit has at least one passive filter. The steam and air mixture is forced through the at least one passive filter due to a differential pressure generated in the fuel handling area. The at least one passive filter traps particulates from the steam and air mixture to produce a filtered steam and air mixture that is released through a second vent mechanism into the atmosphere.
	abstract	A high energy photon source. A pair of plasma pinch electrodes are located in a vacuum chamber. The chamber contains a working gas which includes a noble buffer gas and an active gas chosen to provide a desired spectral line. A pulse power source provides electrical pulses at voltages high enough to create electrical discharges between the electrodes to produce very high temperature, high density plasma pinches in the working gas providing radiation at the spectral line of the active gas. A blast shield positioned just beyond the location of the high density pinch provides a physical barrier which confines the pinch limiting its axial elongation. A small port is provided in the blast shield that permits the radiation but not the plasma to pass through the shield. In a preferred embodiment a surface of the shield facing the plasma is dome-shaped.
	claims	1. A nuclear reactor primary coolant system (10), comprising:a primary coolant pipe (30) which delimits an inner space (32) in which a primary cooling fluid of the nuclear reactor flows, the primary cooling fluid flowing from the upstream direction in a downstream direction of the primary coolant pipe (30), and the primary coolant pipe has a central axis substantially parallel to the primary cooling fluid flow direction;an additional pipe (26) which is tapped from the primary coolant pipe (30), the additional pipe (26) delimiting an inner space which communicates with the inner space (32) of the primary coolant pipe (30);a sleeve (36) which has a first end (50) which is connected to the additional pipe (26), and a second free end (52) which is located in the inner space (32) of the primary coolant pipe (30), the second end of the sleeve opening up toward said central axis;wherein the second end (52) of the sleeve (36) terminates in and is delimited by a free peripheral edge (53) which has at least upstream and downstream sectors (56, 58) which are directed in the upstream and downstream direction of the primary coolant pipe (30), respectively, the upstream sector (56) penetrating more deeply into the inner space (32) from the primary coolant pipe (30) than the downstream sector (58). 2. The system according to claim 1, characterised in that the upstream and downstream sectors (56, 58) penetrate into the inner space (32) over first and second mean penetration depths, respectively, the first depth being greater than the second by at least 10% of the largest dimension of the cross section of the additional pipe (26). 3. The system according to claim 2, characterised in that the first mean penetration depth is greater than 50% of the largest dimension of the cross section of the sleeve (36) in the region of the free end (52) thereof. 4. The system according to any one of claims 1 to 3, characterised in that the peripheral edge (53) has a bevelled profile. 5. The system according to any one of claims 1 to 3, characterised in that the peripheral edge (53) has a notched profile. 6. The system according to claim 1, characterised in that the upstream sector (56) extends over at least 30% of the peripheral edge (53). 7. The system according to claim 1, characterised in that the sleeve (36) comprises a contraction (62) which terminates at the free end (52), the contraction (62) having a smaller flow cross-section than that of the additional pipe (26). 8. The system according to claim 7, characterised in that the contraction (62) extends between a restriction (60) which is formed in the sleeve (36) and the second free end (52), the restriction (60) being arranged in a connection zone of the sleeve (36) to the additional pipe (26). 9. The system according to claim 7, characterised in that the largest dimension of the flow cross-section of the additional pipe (26) is between 1.7 and 3 times the largest dimension of the flow cross-section of the contraction (62). 10. The system according to claim 1, the second end of the sleeve being tubular, wherein the upstream and downstream sectors are part of the tubular end. 11. A nuclear reactor primary coolant system (10), comprising:a primary coolant pipe (30) which delimits an inner space (32) in which a primary cooling fluid of the nuclear reactor flows, the primary cooling fluid flowing from the upstream direction in a downstream direction of the primary coolant pipe (30);an additional pipe (26) which is tapped from the primary coolant pipe (30), the additional pipe (26) delimiting an inner space which communicates with the inner space (32) of the primary coolant pipe (30);a sleeve (36) which has a first end (50) which is connected to the additional pipe (26), and a second free end (52) which is located in the inner space (32) of the primary coolant pipe (30);wherein the second end (52) of the sleeve (36) is tubular and terminates in and is delimited by a free peripheral edge (53) which has at least upstream and downstream sectors (56, 58), the upstream and downstream sectors being part of the second end of the sleeve and being directed in the upstream and downstream direction of the primary coolant pipe (30), respectively, the upstream sector (56) penetrating more deeply into the inner space (32) from the primary coolant pipe (30) than the downstream sector (58). 12. The system according to claim 11, the second end of the sleeve being substantially cylindrical. 13. The system according to claim 11, wherein the free peripheral edge has a closed contour. 14. A nuclear reactor primary coolant system (10), comprising:a primary coolant pipe (30) which delimits an inner space (32) in which a primary cooling fluid of the nuclear reactor flows, the primary cooling fluid flowing from the upstream direction in a downstream direction of the primary coolant pipe (30);an additional pipe (26) which is tapped from the primary coolant pipe (30), the additional pipe (26) delimiting an inner space which communicates with the inner space (32) of the primary coolant pipe (30);a sleeve (36) which has a first end (50) which is connected to the additional pipe (26), and a second free end (52) which is located in the inner space (32) of the primary coolant pipe (30);wherein the second end (52) of the sleeve (36) terminates in an opening delimited by a free peripheral edge (53) having at least upstream and downstream sectors (56, 58) directed in the upstream and downstream direction of the primary coolant pipe (30), respectively, the upstream sector (56) penetrating more deeply into the inner space (32) from the primary coolant pipe (30) than the downstream sector (58).
041892542	summary	The invention relates to a system for the storage of radioactive material in rock cavities. More particularly the invention relates to a repository for the storage of spent fuel from nuclear power plants and such high level waste that is produced during the reprocessing of spent nuclear fuel. Systems for the storage of radioactive material in rock have previously been proposed. Such a system is described in the U.S. patent application Ser. No. 857,041 of Dec. 2, 1977, by Hallenius and Sagefors, and consists of an outer cavity in the rock. This cavity encloses a core of rock mass and is filled with clay so that the clay forms a shell around this core in which is provided an inner cavity which forms the storage space for the radioactive material. This inner cavity is provided with recesses for the accommodation of the radioactive material and communicates with a shaft for entering the radioactive material. This system is provided with an inner cooling system consisting of a plurality of conduits for a coolant. Each such conduit forms a closed loop which extends in a vertical plane along the inside of the inner cavity and along the outside of the core of rock. The system may also be provided with an outer cooling system consisting of a tunnel situated in the rock outside the shell of clay, said tunnel forming a helix which extends concentrically with the system in several turns along the total height of the system. The ends of the helical tunnel are joined in the rock at some distance from the repository, thereby forming a closed cooling system. Therefor, both the inner and the outer cooling system will operate according to the thermosiphon principle which means that coolant heated by the heat developed by the radioactive material rises upwards in the cooling system due to its lower density and is conveyed to a place within or outside the repository having a lower temperature where the coolant is cooled and returned to the hotter places in the repository. Thus, the circulation of the coolant is effected without the aid of any external machinery requiring the supply of energy from outside. However, this cooling system is relatively complicated. Also it is a disadvantage that it is difficult to calculate beforehand the dimensions of the cooling system and the whole repository in order that the dissipation of the generated heat shall be effective without causing dangerous temperature rises in the repository and its environment. For this reason the dimensions of the repository and the cooling system must be estimated with large safety margin which may make the construction costs unnecessarily high. The present invention relates to a repository of the kind described above for the storage of radioactive material in rock. The repository comprises a substantially spherical cavity excavated in the rock. This cavity is surrounded by a shell of rock and this shell is surrounded by a shell of clay. The clay shell is surrounded by the rock formation. It is an object of the invention to provide in a repository of this kind an effective distribution and dissipation of the heat generated by the stored radioactive material. The invention makes it possible to calculate beforehand with great accuracy the heat distribution and the temperature rise in the environment of the repository. Hereby it is also possible to calculate with great accuracy the dimensions of the repository so that the temperature in the rock and the clay shell does not reach dangerous values. The cooling system in the repository according to the invention also contributes to the mechanical stability of the repository and prevents the cavity from collapsing under the action of extremely high external forces. According to the invention the repository is characterized in that a vertically standing tube-shaped member of a heat resistant and mechanically strong material is arranged within the cavity, which tubeshaped member divides the cavity into an outer space and an inner space and is provided at its top and bottom ends with openings connecting the outer space with the inner space, that both the inner space and the outer space are filled with substantially spherical bodies of a heat resistant and mechanically stable material which bodies are provided with through openings and arranged so that these openings extend at an angle to the horizontal plane, and that the radioactive material to be stored is formed into rods which are placed within said openings in some of the spherical bodies in such a way that the rods of radioactive material are at a certain distance from the inside of the openings, and that those of the spherical bodies which contain radioactive material are situated in the bottom part of the inner space in the tubeshaped member. The tubeshaped member preferably consists of a cylindrical tube of reinforced concrete which is open at both ends and also is provided with openings around its periphery adjacent to its ends. The said spherical bodies are also preferably made of reinforced concrete. In the repository according to the invention air in the bottom part of the tubeshaped member will be heated by the radioactive material and caused to rise upwards within the tubeshaped member to its top end where the air is forced through the openings at the top end against the wall of the cavity where the air is cooled and flows downwards in the outer space between the tubeshaped member and the wall of the cavity, whereupon the air again flows into the tubeshaped member through the openings at its bottom end and again comes in contact with the radioactive material and is heated anew so that the flow cycle is repeated. The air flows through the spaces between the spherical bodies and through the openings in these bodies. Thus, the spherical bodies act as a porous mass which makes possible a relatively free and rapid air flow and simultaneously prevents the cavity from being compressed and collapsing under the action of high external forces. The heat generated by the radioactive material is thus distributed by convection nearly uniformly over the whole cavity, and large temperature peaks in limited areas of the interior of the cavity are avoided. The generated heat spreads through the rock surrounding the cavity and further on to the clay shell. Due to the spherical shape of the cavity it is relatively simple to calculate the temperature distribution in the environment of the cavity. For a given amount of stored radioactive material it is thus possible to estimate the variation with time of the temperature in the rock and the clay shell and the resulting maximum temperatures. These temperatures will of course be dependent of the dimensions of the rock mass and the clay shell, and it is therefore possible to determine beforehand these dimensions so that the temperature cannot assume critical values. By "critical values" of the temperature are meant such values which may cause undesirable changes in the rock and the clay, e.g. crumbling of the rock and drying-up of the clay so that it loses its plasticity.
	summary
060375175	summary	TECHNICAL FIELD OF THE INVENTION This invention relates to waste treatment systems, and more particularly to an apparatus and method for treating waste materials which include beta radiation emitting constituents. BACKGROUND OF THE INVENTION Radioactive waste materials are generated at a number of different sources, including nuclear power plants, nuclear weapons facilities, and nuclear fuel processing facilities. There are also a number of less obvious sources of radioactive wastes. For example, the fly ash of coal power plants may include radioactive constituents. Also, radioactive materials are used in certain medical procedures. Thus, medical facilities are major producers of radioactive waste materials. The radioactive wastes produced at medical facilities include equipment and clothing which may be contaminated by the radioactive material used in medical procedures. The three types of radiation emitted from radioactive materials are alpha, beta, and gamma radiation. Alpha and beta radiation comprise particles which are emitted from the nucleus of an atom, while gamma radiation comprises short-wavelength photons of nuclear origin. Alpha particles are doubly ionized helium nuclei, and thus have a net positive electrical charge. Beta radiation comprises primarily electrons, although some radioactive isotopes emit positrons which are also referred to as beta particles. Both the charged alpha particles and beta particles may be deflected by an electromagnetic field, although beta particles are deflected more easily due to their lower mass. Gamma radiation is either emitted from a radioactive material directly or emitted as the result of a collision between an alpha or beta radiation particle and some other particle. Radioactive materials may emit one or more of the three different types of radiation, alpha, beta, of gamma radiation. Many radioactive materials emit primarily only alpha particles and/or beta particles, but produce gamma radiation indirectly as the high-energy alpha and beta particles collide with other particles. Wastes which include radioactive materials may be treated in a molten metal process to remove organic materials and to tie up the radioactive material. U.S. patent application Ser. No. 09/096,617, filed Jun. 12, 1998 by the present inventor, discloses an apparatus and method for treating waste streams which include radioactive constituents. The apparatus and method disclosed in Application Ser. No. 09/096,617, which is incorporated herein by reference, removes organic constituents from the mixed waste stream and contains the radioactive constituents. Organic materials in the waste stream react with the molten reactant metal to produce primarily elemental carbon, hydrogen, nitrogen, and metal salts. Radioactive materials in the waste stream are alloyed in the molten metal for eventual storage. The molten metal process disclosed in Application Ser. No. 09/096,617, utilizes radiation absorbing metals such as lead and tungsten, for example, in the molten reactant metal in order to safely absorb radioactive emissions from the alloyed radioactive materials. SUMMARY OF THE INVENTION It is an object of the invention to provide an apparatus and method for treating wastes which include radioactive constituents, particularly beta radiation emitters. The apparatus according to the invention includes a molten metal reactor and an electromagnetic field generating arrangement. The molten metal in the reactor reacts with any organic constituents in the waste material, and alloys the radioactive constituents. The electromagnetic field generating arrangement produces a unidirectional electromagnetic field extending through the molten metal and through at least one target area preferably within the molten metal. This unidirectional electromagnetic field directs or deflects beta radiation toward the target area and into a replaceable radiation absorbing module positioned in the target area. It is also believed that the intense electromagnetic field may enhance the beta emissions. In any event, radiation absorbing material included in the module absorbs the beta radiation in a stable form. The radiation absorbing material in the module also absorbs gamma radiation produced as beta particles and alpha particles are absorbed by the radiation absorbing material. The molten metal reactor includes a reactor chamber charged with a suitable reactant metal. A heating arrangement is included in the molten metal reactor for heating the reactant metal and maintaining the reactant metal in a molten state at a desired reaction temperature. A waste input structure is preferably included for introducing the waste material into the reactor in position for a submerging arrangement to submerge the material in the molten reactant metal. A circulating arrangement may be included for circulating the molten metal within the reaction vessel and ensuring that the radioactive constituents circulate through the area of the molten metal traversed by the electromagnetic field. The preferred molten metal reactor also includes an arrangement for removing reaction products from the molten metal reactor and also an arrangement for adding additional reactant metal. The reactant metal used in the molten metal reactor may comprise any suitable reactant metal. The primary constituent of the reactant metal preferably comprises aluminum although magnesium and/or lithium may be used with or instead of aluminum. In one preferred form of the invention, the molten metal comprises primarily aluminum along with lesser fractions of other constituents such as iron, copper, zinc, and calcium, for example. The reactant metal also preferably includes one or more radiation absorbing metals such as lead, tungsten, palladium, cadmium, dysprosium, and europium. The electromagnetic field generating arrangement comprises at least one coil of electrically conductive material. The coil or coils are preferably located within the molten metal in position to produce a highly focused electromagnetic filed in at least one target area. Each coil is encased within a protective material to protect the coil material from reacting or alloying with the molten metal. In the preferred form of invention, each coil is made from a tubular conductor material such as copper. The invention includes a cooling system comprising a coolant fluid supply and pump for circulating the coolant fluid through each tubular conductor and cooling the conductor material. A heat exchanger may be used for cooling the coolant fluid prior to returning the fluid to the coolant supply. The electromagnetic field produced through the molten reactant metal according to the invention is unidirectional, that is, the field does not alternate directions. The direction of the field is such that it directs beta particles to the target area or areas and thus to the radiation absorbing modules positioned in each target area. The field generating arrangement includes a voltage supply for inducing a current through each coil to produce the desired field. In the preferred form of the invention the voltage is pulsed to produce a pulsed unidirectional electromagnetic field. The pulsed electromagnetic field creates a pumping or circulating action within the molten reactant metal. Each radiation absorbing module includes a material suitable for absorbing the radiation which is directed to the target area by the electromagnetic field. In the preferred form of the invention, each module includes alternating layers of tungsten and lead with a spacing arrangement to maintain the spacing between tungsten layers. The spacing arrangement may comprise tungsten spacer extensions formed on each tungsten layer. In any case, all of the radiation absorbing material in each module is preferably encased in a protective material. The protective material protects the radiation absorbing material so that the module may be positioned in contact with the molten reactant metal without losing the radiation absorbing metals to the melt. In the preferred form of invention, one radiation absorbing module is mounted on a positioning structure which allows the module to be positioned in a target area within the molten reactant metal. The preferred target area is an area at which the electromagnetic field strength is greatest. One or more additional radiation absorbing modules may also be positioned in different target areas traversed by the electromagnetic field. The apparatus according to the invention directs beta radiation emissions from radioactive waste materials in the molten reactant metal to the replaceable radiation absorbing modules. Thus, beta radiation may be absorbed in the module without allowing the radioactive materials to commingle and alloy with the radiation absorbing material of the module. Absorbing the beta radiation with these replaceable, isolated modules effectively increases the capacity of the system to handle beta emitting materials without increasing the volume of the reactant metal, and thus the volume of material which is then contaminated with the radioactive material. These and other objects, advantages, and features of the invention will be apparent from the following description of the preferred embodiments, considered along with the accompanying drawings.
	summary
	claims	1. A method for automatic identification of execution phases in load test data, comprising:receiving load test data indicating processor utilization for a plurality of threads over a period of time;dividing the period of time of the load test data into a plurality of intervals;for each pair of proximate intervals of the plurality of intervals, determining whether a statistical characterization of thread-wise processor utilization for a first interval of the pair of intervals is statistically indistinguishable from a statistical characterization of thread-wise processor utilization for a second interval of the pair of intervals;combining the pair of proximate intervals into a single interval when it is determined that the statistical characterization of processor utilization for the first interval is statistically indistinguishable from the statistical characterization of processor utilization for the second interval, for each of the plurality of threads; anddividing each of the pair of proximate intervals into subintervals when it is determined that the statistical characterization of processor utilization for the first interval is not statistically indistinguishable from the statistical characterization of processor utilization for the second interval, for at least one of the plurality of threads, wherein one or more execution phases are automatically identified as occurring between proximate intervals that are not substantially equivalent,wherein the steps of receiving the load test data, dividing the period of time, determining indistinguishablity, combining proximate intervals, and dividing proximate intervals are performed using one or more computer systems. 2. The method of claim 1, wherein the processor is a central processing unit (CPU). 3. The method of claim 2, wherein the thread-wise statistical characterization of processor utilization is a mean CPU utilization for each of the plurality of threads. 4. The method of claim 2, wherein the thread-wise statistical characterization of processor utilization is a standard deviation or variance of CPU utilization for each of the plurality of threads. 5. The method of claim 4, wherein determining whether the mean CPU utilization for the first interval is statistically indistinguishable from the mean CPU utilization for the second interval includes performing a modified Student's T test. 6. The method of claim 5, wherein determining whether the mean CPU utilization for the first interval is statistically indistinguishable from the mean CPU utilization for the second interval includes performing Welch's modification of Student's T test with unequal variances and unequal sample sizes. 7. The method of claim 1, wherein the period of time of the load test data is initially divided into a plurality of intervals of equal duration prior to performance of the steps of combining and dividing. 8. The method of claim 1, wherein the process of dividing intervals into subintervals and comparing subintervals is performed recursively up to a desired level of granularity. 9. The method of claim 1, wherein the determination as to whether the statistical characterization of thread-wise processor utilization of the pair of intervals is statistically indistinguishable is performed using a predetermined confidence interval. 10. The method of claim 1, wherein the determination as to whether the statistical characterization of thread-wise processor utilization of the pair of intervals is statistically indistinguishable is performed using multiple different confidence intervals, with each of the multiple different confidence intervals applied to calculating a statistical characterization of processor utilization for a different thread. 11. The method of claim 1, additionally including using the identified execution phases to correlate execution phases with application code segments associated with the load test data to identify application code segments that are responsible for phases of relatively high thread-wise processor utilization. 12. A method for automatic identification of bottlenecks n application code, comprising:executing the application code and recording load test data indicating CPU utilization for a plurality of threads over a period of time:dividing the period of time of the load test data into a plurality of intervals;for each pair of proximate intervals of the plurality of intervals, determining whether a statistical characterization of thread-wise CPU utilization for a first interval of the pair of intervals is statistically indistinguishable from a statistical characterization of thread-wise CPU utilization for a second interval of the pair of intervals;combining the pair of proximate intervals into a single interval when it is determined that the statistical characterization of CPU utilization for the first interval is statistically indistinguishable from the statistical characterization of CPU utilization for the second interval;dividing each of the pair of proximate intervals into subintervals when it is determined that the statistical characterization of CPU utilization for the first interval is not statistically indistinguishable from the statistical characterization of CPU utilization for the second interval, wherein one or more execution phases are automatically identified as occurring between proximate intervals that are not substantially equivalent; andusing the identified execution phases to correlate execution phases with segments of the application code associated with the load test data to identify segments of the application code that are responsible for phases of relatively high thread-wise CPU utilization,wherein the steps of executing the application code and recording load test data, dividing the period of time, determining indistinguishability, and correlating execution phases are performed using one or more computer systems. 13. The method of claim 12, wherein it is determined that the statistical characterization of CPU utilization for the first interval is statistically indistinguishable from the statistical characterization of CPU utilization for the second interval when the statistical characterization is statistically indistinguishable for every thread of the plurality of threads. 14. The method of claim 12, wherein it is determined that the statistical characterization of CPU utilization for the first interval is not statistically indistinguishable from the statistical characterization of CPU utilization for the second interval when the statistical characterization is not statistically indistinguishable for at least one thread of the plurality of threads. 15. The method of claim 12, wherein the thread-wise statistical characterization of CPU utilization is a mean CPU utilization for each of the plurality of threads. 16. The method of claim 12, wherein the thread-wise statistical characterization of CPU utilization is a standard deviation or variance of CPU utilization for each of the plurality of threads. 17. The method of claim 12, wherein determining whether the mean CPU utilization for the first interval is statistically indistinguishable from the mean CPU utilization for the second interval includes performing a modified Student's T test. 18. The method of claim 17, wherein determining whether the mean CPU utilization for the first interval is statistically indistinguishable from the mean CPU utilization for the second interval includes performing Welch's modification of Student's T test with unequal variances and unequal sample sizes. 19. A computer system comprising:a processor; anda non-transitory, tangible, program storage medium, readable by the computer system, embodying a program of instructions executable by the processor to perform method steps for automatic identification of execution phases in load test data, the method comprising:receiving load test data indicating CPU utilization for a plurality of threads over a period of time;dividing the period of time of the load test data into a plurality of intervals;for each pair of proximate e intervals of the plurality of intervals, determining whether a statistical characterization of thread-wise CPU utilization for a first interval of the pair of intervals is statistically indistinguishable from a statistical characterization of thread-wise CPU utilization for a second interval of the pair of intervals based on performing Welch's modification of Student's T test with unequal variances and unequal sample sizes;combining the pair of proximate intervals into a single interval when it is determined that the statistical characterization of CPU utilization for the first interval is statistically indistinguishable from the statistical characterization of CPU utilization for the second interval, for each of the plurality of threads; anddividing each of the pair of proximate intervals into subintervals when it is determined that the statistical characterization of CPU utilization for the first interval is not statistically indistinguishable from the statistical characterization of CPU utilization for the second interval, for at least one of the plurality of threads, wherein one or more execution phases are automatically identified as occurring between proximate intervals that are not substantially equivalent. 20. The computer system of claim 19, wherein the determination as to whether the statistical characterization of thread-wise CPU utilization of the pair of intervals is statistically indistinguishable is performed using multiple different confidence intervals, with each of the multiple different confidence intervals applied to calculating a statistical characterization of CPU utilization for a different thread. 21. The computer system of claim 19, wherein additionally including using the identified execution phases to correlate execution phases with application code segments associated with the load test data to identify application code segments that are responsible for phases of relatively high thread-wise CPU utilization.
	description	The present invention was made under Contract No. W911QX-04-C-0097, mod. no. P00001, awarded by the Defense Advanced Research Projects Agency and the United States Government has certain rights in this invention. The invention relates to an electrostatic actuator apparatus and method, and more particularly, to a beta emission process of a source material emitting electrons which are then captured by a target material wherein electrical work is performed which in turn is transferred into mechanical work in the form of rotation of a rotor. Specific applications include a radioisotope fueled rotary actuator for micro and nano air vehicles employed as the main form of propulsion. A nano air vehicle (NAV) is commonly defined as an air vehicle with a maximum dimension of 7.5 centimeters in any axis, weighing 10 grams or less, and capable of at least 20 minutes endurance and 1-kilometer range. At this scale, the aerodynamics and power are significant challenges. Multiple tradeoffs are involved. One approach to the challenges is to use lithium polymer batteries as part of the airframe. However, these tend to change size as they are used, impacting structure integrity. Another challenge is motor integration. Conventional fossil or battery powered motors are driven through gear mechanisms to turn a propeller or turbine. These designs impose severe range penalties, have a high specific mass fraction of total air vehicle, and endurance limitations which impact the overall ability to perform specific missions. Propulsion of micro air vehicle (MAV) and NAV aircraft is traditionally affected through a conventional motor. The problem is that a motor, either electrical (usually DC) or fossil fueled, occupies a rather large portion of the vehicle mass fraction. Furthermore, as the MAV/NAV vehicle becomes more structurally efficient, the percentage devoted to propulsion utilizing conventional technologies increases further, implying an even harsher penalty in terms of payload due to propulsion requirements. Another significant drawback to conventional propulsion technology is implicit in the range capability of the vehicles. As the vehicles shrink in size, the ability to carry fuel, either electrical or fossil based, enforces strict limits on overall endurance. Thus, a vehicle which is capable of flight at just 30 knots is typically only capable of ranges 7-10 nautical miles. While this may be suitable for some applications, under true operational constraints, a significant increase in range is required in order to make the technology truly valuable to the user community. Endurance is another area requiring improvement for MAV/NAV devices. Present technological limits are in the 30 minute range as evident by the AeroVironment® Black Widow design, typical of high performance MAV's. The flight duration under optimal conditions significantly limits utility to the operator. AeroVironment is a registered trademark of the AeroVironment Inc. Corporation of California. Radioisotope power systems (RPS) are employed in spacecraft. Radioisotope thermoelectric generators (RTGs) have been used to power, for example, pacemakers and spacecraft, but are complex, requiring the source material, a walled container, thermocouples, and a heat sink to generate electricity. Nor is the energy produced by these compatible with the mass and volume constraints of MAV/NAVs. What is needed, therefore, are techniques for an actuator for micro-scale vehicles that is efficient in power conversion, providing sufficient power to generate lift and thrust at this small scale of flight. Embodiments significantly reduce the mass fraction devoted to air vehicle propulsion and increase range and operation time for MAV/NAV vehicles. Volumetric and electrical efficiencies are maximized to obtain an operational duration of 1 year for a propeller driven MAV design. The propulsion technique is directly applicable to all other types of micro and nano scale vehicles including ground and water conveyances, including submersibles. Embodiment applications encompass new robotic devices including home products. Through the use of radioisotopes as the fuel and designing the rotary actuator to take advantage of the intrinsically high operating voltage and resulting high rotational frequency, direct drive is possible in embodiments. This eliminates the need for gearing and ancillary fuel storage containers. Therefore, the mass fraction devoted to propulsion for the vehicle can be significantly altered such that an increase in payload is permissible. This increases the utility of the overall vehicle concept. Furthermore, by using a radioisotope material with a sufficiently long half life, the endurance and therefore the range can be significantly improved allowing the vehicle to fly great distances and loiter over the mission area for weeks to potentially months before returning to the home base. This is a more extensive hover and stare capability than currently available. For embodiments, the radioisotope provides not only propulsive power but electrical power. Embodiments of the present invention utilize a rotary actuator as a means of driving a conventional propeller for propulsion of the MAV/NAV. The use of a radioisotope powers an EA at the MAV/NAV scale at the power density level of ≧100 mW/gm. Embodiments are scalable over a range of absolute powers and corresponding sizes. Furthermore, by using a radioisotope to drive the actuator two significant improvements are enabled which dramatically change the MAV/NAV paradigm. The first is a reduction in overall mass fraction of the air vehicle devoted to propulsion and the second is virtually unlimited range and endurance on the order of three months to a year in embodiments. Furthermore, this technology is not limited to conventional air vehicle designs. The rotary actuator, through a cam and follower motion transmission design, can also drive unconventional flapping wing designs such as an ornithopter. Embodiments can also mimic the biomechanics of insect type devices to effect flight in a manner analogous to dragonflies and hummingbirds. As mentioned, additional applications include propulsion for micro-submersible vehicles wherein the radioisotope provides not only propulsive power but electrical power to the vehicle. By virtue of elimination of combustion, virtually unlimited range is available. Applications would not suffer from radiation emission safety hazards and so could avail themselves of more energetic sources. Other applications are NASA projects involving miniature robotic payloads which are keenly attuned to the needs for extremely low weight (during launch) and the desire for extremely long operation which is not necessarily afforded through conventional battery technology and solar charging circuits. Additional space applications include robotic devices which “fly” in an unconventional sense over a planetary surface and thus require both propulsion and extremely low weight. One embodiment provides a rotary electrostatic actuator (EA) apparatus comprising a high voltage source; a target material receiving voltage from the high voltage source; wherein a source vane is attracted to the target material as a result of charges attracted to higher E fields. Another embodiment comprises a radioisotope emission high voltage source. Further embodiments provide at least one of piezoelectric crystals and Van de Graff generator. Yet other embodiments comprise S35; P32; P33; Ca45; and Sn123. In yet further embodiments, the actuator is a disk rotor; a vertical wall rotor; and a stacked rotor. Some embodiments comprise a replaceable source. An embodiment implements partial discharge. Another embodiment comprises a storage capacitor re-charging the target. Another embodiment is an electrostatic rotary actuator method comprising providing emission from a source; capturing the emission by a target material; generating rotation from electrostatic force; and discharging developed potential. In another embodiment, the source is a radioisotope providing the emission. For other embodiments, the step of discharging comprises partial discharge. Additional embodiments provide a low atomic number beta emitter source. In a yet further embodiment, the radioisotope source further provides electrical power. Embodiments include a radioisotope fueled electrostatic disk rotary actuator nano air vehicle apparatus comprising two pairs of chutes comprised of metal, wherein the two pairs of chutes comprise a surface film of a light metallic element; a rotating vane disk, between the two pairs of chutes and coaxial with the two pairs of chutes, the rotating vane disk comprising twenty four source vanes comprising beta-emitting radioisotope comprising at least one of S35 and Ca45, wherein the radioisotope comprises a source film with a thickness of about approximately one half penetration depth, whereby current is a maximum; a housing comprising a lead-plated vacuum envelope, enclosing the two pairs of chutes and the rotating vane disk, whereby emission products of the radioisotope are contained, the vacuum envelope is sputtered deposition plated with a lead layer of about approximately one micron, whereby surrounding area is protected from soft X-rays, and beta upset of localized electronics is prevented; and wherein rotation of the rotary actuator is magnetically coupled directly to a propeller component, thereby eliminating losses due to a mechanical gear box, whereby propulsion is provided to the nano air vehicle. The features and advantages described herein are not all-inclusive and, in particular, many additional features and advantages will be apparent to one of ordinary skill in the art in view of the drawings, specification, and claims. Moreover, it should be noted that the language used in the specification has been principally selected for readability and instructional purposes, and not to limit the scope of the inventive subject matter. Overview An actuator's source vanes rotate within an electric field between chutes' walls, generating torque. The principal which allows torque and power is the change in energy as a vane gets closer to the outer walls. The general equation is Torque=d(Energy)/d(theta)=d(½ C V^2)/d(theta). Hence, the vane is attracted toward the narrow sections of the “chutes”. The energy is proportional to the volume of the actuator, so large actuators will have more power and torque. Also, energy is proportional to the E field squared, hence, in embodiments, the E field is held close to the breakdown E field. The power is torque frequency, so faster charging times and greater rotor RPM generate more power, provided that the electrical current can charge the plates at the higher frequency. Rotor actuator embodiments provide an electrical-to-mechanical power conversion, and the upper limit of the mechanical power is a fraction of the electrical power, which is VoltageCurrent. The larger the current, the more power is generated. For radioisotopes, this means a larger source disk will generate more power. Also, for minimum size, in embodiments, the gap dimension is slightly larger than the threshold distance necessary to prohibit breakdown. The rotor can be scaled in size for more power. Due to electrostatic breakdown (arcing/lightning), the gap between the vane and the chute of the rotor will scale as the voltage is increased, and the E field will be a constant. For example, in embodiments, the peak E field is at a value just below the threshold for arcing. As stated, the power of the actuator scales as E field squared, so it is beneficial to maintain high E fields. A rotary actuator can have reasonably arbitrary spin RPM, unlimited by mechanical resonant frequencies, and hence can generate power just limited by the ability of the current to charge the plates. The rotor can be charged by anything that generates high voltage. Examples are piezoelectric crystals or Van-de-Graff generators. It is not required that radioisotopes, embedded into the spinning disk, be the energy source. The charging source can be separate from the actuator. For embodiments, the number of source “wedges” is much larger than the number of chutes, so that, as one wedge is getting discharged at the throat (end) of the chute, the torque is not disturbed for the next wedge. The wedge getting discharged is effectively at the chute voltage, and hence the next wedge will be attracted to this discharged wedge, which will not generate torque, because the discharged wedge is on the same disk. For this reason, in embodiments, there is a throat section at the end of each chute, which fully encapsulates the wedge getting discharged. Hence, the next-in-line wedge always is closer to the chute walls during the discharge process, and not closer to the discharging wedge. Partial discharge can be used as a technique to keep the same power level at arbitrary RPM. With full discharge, the RPM is limited by the time is takes to re-charge the vanes. With partial discharge, smaller currents are not required to waste time and energy charging up the vane/capacitor to close to peak voltage. Almost all the torque is generated when the capacitance is near fully charged, due to the E^2 dependence of the torque. Partial discharge can be implemented as a long RC time constant, when the spinning “wedge” of the disk is shorted at the end of the chutes. It can also be implemented as a separate storage capacitor, which quickly re-charges the chute. Disk rotor embodiments with a flat spinning disk, have the advantage that they are mechanically robust. The flat spinning disk does not experience twist. However, the charge closer to the inner radius is not as effectively generating torque as the charge on the outer radius of the disk. The charge on the inner radius of the disk is also more susceptible to arcing. Vertical wall embodiments have the advantage of more torque, due to all the force being exerted at the farthest radius of the rotor. They are less susceptible to arcing, because the vertical surface of the rotor sees the same large dimensions. They also have the advantage that the outer vertical walls of the disk can be tall and have more surface area and hence more torque. However, mechanically, they are a more 3-dimensional structure that needs to be more rigid due to the strain on the outer walls of the spinning disk. Using radioisotopes, a vacuum is employed in embodiments to allow the electrons or alpha particles to cross the gap between the spinning disk and the chutes. Without a vacuum, electrons ionize the intervening gas, not charging the capacitance. As shown by the standard “Paschen Curve”, the pressure, for millimeter size gaps or larger, is much less than 1/1000th of an atmosphere. Pressures above 1/1000 atmosphere stop charge carriers. Pressures near 1/1000 atmosphere cause cascading ionization (lightning). This vacuum constraint is not necessary, for example, if embedded radioisotopes are not used to charge the vanes, and instead an external source is used. Using radioisotopes, about one third of the charge carriers make it across the gap and charge the plates. Since radioisotopes emit charged particles in random directions, only one third have enough energy in the gap direction to bridge the high voltage across the gap, contributing to efficiency considerations. Electrostatic actuator design particulars follow. Beta emission characteristics are explained as power modes. For example, operation life is highly dependent upon isotope selection. Embodiments provide weeks to months of useful life. Some embodiments may significantly extend this parameter. Rechargeable designs are included. For embodiments where only beta emission is employed, candidate isotopes are considered. X-ray emission can be a concern for some embodiments. However, judicious choice of materials can limit this to manageable levels. Aluminum structure, for example, generates softer X-ray spectrum than copper. Shielding is possible without significant impact to power density for “light” materials. For embodiments, the beta emitter has the lowest atomic number with acceptable half life which decays into stable elements, or maintains beta decay throughout the process. For some embodiments, this eliminates materials such as Ru106 which decays into gamma emitter. For embodiments, Sn123 is a gamma emitter which meets operational life and power density requirements. Embodiments demonstrate size, power density, frequency, absolute power, and stroke to support NAV size scale applications. Figure Details FIG. 1 depicts four rotary actuator embodiments 100 configured in accordance with the present invention. The rotary motor actuator design operates at a lower voltage than a parallel plate approach; however, embodiments require a vacuum envelope. In addition, embodiments employ either a bearing or very low friction bushing to benefit from the high rotational velocities. High rotational frequencies and variable rates support MAV/NAV operation. Actuator embodiments are sized over a power (VI) range of 0.005<=P<=3.8 W. Embodiments are scalable over a large range of powers/size. TABLE 1E fieldVolt.VIMech.MechLargestSmallestmaxMaxpowerPowerPower/MassGapGapCase(V/m)(kV)(W)(W)mW/gm(mm)(mm)1) Stepped5.50E+071100.1400.01638.04.02.0ParallelPlate2) 1 vane,5.50E+07160.0050.0025.44.00.21 spiral3) 24 vane,5.50E+07400.1500.080188.04.00.41 spiral4) 24 vane,1.00E+08843.8000.8702046.02.00.44 spiral4) 24 vane,5.50E+07421.1000.254598.02.00.44 spiral The embodiments depicted in FIG. 1 (first through fourth rotor embodiments in FIGS. 1A-1D, respectively) and the table above have mechanical powers calculated for two different E fields: 5.5e7 and 1.0e8 V/m (the second E field is double the first E field, and the voltage is doubled as well). In embodiments, the Power is four times larger when the voltage is doubled, for any of these actuators, because Power ˜V^2. The two E fields were selected assuming that one of these would be an upper bound below which arcing occurs. Embodiment case 4 (FIG. 1, FIG. 1A disk rotor embodiment one), is a noted case for comparison. It is the rotary actuator using 4 chutes and 24 source wedges. It achieves, at a large E field, a Power/Mass of 2048 mW/gm. For rotary embodiments, there is a limit to the number of source wedges that can be created around the source disk. The limit is caused by arcing considerations. At the throat of each chute, where each source wedge is individually discharged, there is now a voltage between the discharged source wedge and the neighboring un-discharged source wedge. Hence, there is a large E field, and arcing considerations apply just as between the source wedges and the metal chutes. In embodiments, this gap between source wedges may be free from material so that it is not conductive due to ionization. An advantage of rotary embodiments is that they can operate at 500 Hz with no mechanical decelerations to overcome. However, a bearing needs to be used in embodiments. For embodiments, this bearing can be contained within a vacuum package, and hence itself not require a vacuum seal. For embodiment case 1 (FIG. 1, FIG. 1D rotary embodiment four) and case 2 (FIG. 1, FIG. 1C rotary embodiment three), the main torque during a full cycle occurs when the tip of the single source wedge gets close to the throat of the chute. This only occurs once each cycle when only one large source wedge is used. Embodiment case 3 (FIG. 1, FIG. 1B rotary embodiment two) overcomes this situation of case 2 by breaking the large single source wedge into 24 separate smaller source wedges. Now, the main torque is occurring 24 times each cycle. Case 2 also has another consideration. Torque is proportional to the change in the capacitance between the source wedge and metal chute as a function of angle. By using one large chute, this slope in capacitance is very small over the first 70% of the cycle. Hence, case 4 has 4 chutes to have this large slope region of the chute occur over a larger fraction of the cycle. Finite element modeling (FEM) produced some results for capacitance v. rotation for 1 cm diameter embodiments. 1) More power is derived from more chutes. But an objective is to allow voltage to be high during the greatest slope of the capacitance. Hence, embodiments can not have too many chutes. More power is achieved with more chutes, to a limit. First, the source wedges need to have enough source film to be able to fully charge each chute as it passing through it. Second the number of chutes should probably be half the number of source wedges, in order for embodiments to optimize the torque on each source wedge. 2) The spiral smooth taper has a larger slope to the capacitance (more torque) compared to steps in a parallel plate design. The spiral chute embodiment has a larger change in the capacitance as a function of angle of the wedge source. The spiral also has less sharp corners to cause arcing. 3) For embodiments, it is better to have a least twice as many source pieces than the number of chutes. This allows a steady torque to be applied, and also allows more total capacitance. One source wedge can be experiencing the most torque near the throat of the chute, and the other source wedge(s) can be in the charging stage. Basically, in embodiments, one does not get double the torque by having double the width of the source wedge, because only the leading edge of the source wedge is experiencing the most torque. FIG. 2 depicts perspective views of an additional four rotary actuator embodiments 200 configured in accordance with the present invention. They include rotary embodiment five FIG. 2A, stacked rotor embodiment six FIG. 2B, rotary embodiment seven FIG. 2C, and vertical wall rotor embodiment FIG. 2D. Due to the high collection voltage for embodiments of FIGS. 1 and 2 (which is a function of the nuclear emission process); each unit is encased in a vacuum envelope. In embodiments, the vacuum envelope provides a secondary function in that it is plated, using a sputtered deposition process, with an approximately micron thick film of protective material such as lead to prevent accidental contamination of the surrounding area by soft X-rays and prevent beta upset of any localized control electronics. FIG. 3 is a perspective view 300 of disk rotor embodiment one, FIG. 1A. References include Z-axis 305 and Y-axis 310. Vane 315 is between upper chute pair 320 and lower chute pair 325 and shares Z-axis 305 with them. In embodiments, source material is on both the top and bottom of the source disk, in order to get current flowing in both directions. In alternate embodiments, a thick disk of source material is sandwiched inside a thin walled disk, and beta electrons escape from both sides of the disk. For other embodiments, the source disk is removable to allow for recharging. FIG. 4 is a plan top view 400 of vanes of disk rotor embodiment one, FIG. 1A. Here, references include X-axis 405 and Y-axis 410. Vane disk 415 has center portion 420. FIG. 5 is a side elevation view 500 of disk rotor embodiment one, FIG. 1A. References include Z-axis 505. Vane disk 510 is between upper chute pair 515 and lower chute pair 520 and shares Z-axis 505 with them. Vane disk 510 has center portion 525. For embodiments, the highest gap is 2 mm, there are 24 source vanes, and four metal chutes. FIG. 6 depicts views 600 of rotary embodiment six configured in accordance with the present invention. FIG. 6A is a side view and FIG. 6B is a top view. FIG. 7 depicts views 700 of rotary embodiment five and vertical wall rotor embodiment configured in accordance with the present invention. FIG. 7A is a top view of rotary embodiment five of FIG. 2A, and FIG. 7B is a top view of vertical wall rotor embodiment of FIG. 2D. FIG. 8 depicts capacitance versus rotation relationships diagrams 800 configured in accordance with the present invention. FIG. 8A depicts an angle of −85 degrees, FIG. 8B depicts an angle of 40 degrees, and FIG. 8C depicts an angle of 0 degrees. Relationship equations include: τ = 1 2 ⁢ ⅆ C ⅆ θ ⁢ V 2 Eq . ⁢ 1 P = τω Eq . ⁢ 2 For embodiments, torque and power are optimized by optimizing the slope of C, and allowing the voltage to be large. Just as Force=(½) (dC/dx) V^2, the torque has the same type of formula: Torque=(½) (dC/d(angle)) V^2. Power for linear oscillations is Power=Forcevelocity. Power for spinning motion is Power=Torque(angular velocity). FIG. 9 depicts finite element model (FEM) 900 of rotary embodiment one (FIG. 1A). The FEM model high-frequency structure simulation (HFSS) is used to calculate the capacitance versus rotation angle of each source vane, including probe port to measure impedance between source plate and target. It is a FEM model for a 1 cm diameter rotary motor embodiment. Variables include a 2 mm highest gap, 24 source vanes, and 4 metal chutes. The finite element model models capacitance between source wedge 905 and the metal chute as a function of angle of the source wedge, and includes probe port to measure impedance between source plate and target chutes 910. The impedance near the origin in the figures was determined, and the capacitance was determined from this impedance, using Capacitance=1/(2 pi freq impedance). A low enough frequency was used in the model such that the wavelength was much larger than the actuator (1 MHz), and the impedance should be purely capacitive. The symmetry plane imposes the condition that all the fields are parallel to this plane. The far walls and ceiling of the large cylinder are declared as a radiation boundary. FIG. 10 depicts FEM capacitance predictions 1000 of rotary embodiment one (FIG. 1A). Again, FEM capacitance predictions are for a 1 cm diameter rotary motor embodiment. The slope of the capacitance versus angle is notably important when the source material is deep inside the chute and the voltage is very large. In the plot, that condition occurs from −60 to −90 degrees 1005. The steeper slope in the capacitance between 0 and −20 degrees 1010 occurs just after the source wedge was stripped of its charge by sliding through a slot. The charge buildup process has started but there is little voltage buildup yet. FIG. 11 depicts capacitance vs. rotation 1100 of rotary embodiment one (FIG. 1A). Results are for capacitance vs rotation, 1 cm diameter rotary motor embodiment; total power using 24 source vanes at 100 Hz; 0.250 Watts Total Power/Mass=598 mW/gm; Power VI=1.1 Watts; Largest Voltage=44 kV; Largest E field=5.5e7 V/m; current per source vane=2e-6 Amps; Approximately 23% VI to mechanical power conversion. In plot FIG. 11A for the torque versus rotation angle, the torque only becomes prominent during the last ⅓rd of the cycle (from 60 to 90 degrees). This is when the voltage has reached a maximum, and when the slope of the capacitance is large. In plot FIG. 11B for voltage versus rotation angle, the voltage charges up from 0 to 60 degrees, and peaks between 60 and 90 degrees. The voltage drops from 80 to 90 degrees because the gap in decreasing, but the E field (plot FIG. 11C) is still increasing. In this embodiment example, the E field was limited to 5.5e7 V/m. Nearly four times more power is possible if the E field is allowed to rise to 1e8 V/m. FIG. 12 depicts radioisotope energy analysis diagram 1200 configured in accordance with one embodiment of the present invention. Included is perspective view FIG. 12A of collector plates and force plates with charge collector plates 1205, force plates (less gap) 1210, and depicting wider tip 1215 to get shorting to target plates only at one instant. In addition to perspective view are side view FIG. 12B and top view FIG. 12C. Analyses were for radioisotopes using energy arguments. This particular embodiment is good for explanation purposes. The following conceptual comparisons between four radioisotopes are based on energy arguments, and not on any one particular geometric shape for the rotary motor. Assume an optimal shape is determined, and assume that ¼ the electrical power is converted into mechanical power. Now compare the different radioisotopes, and the power limits for each. FEM can be used to optimize the complex capacitance versus rotation angle for embodiments. Maximize the number of chutes as the radioisotope current can charge in the shorter time frame. More radioisotope source wedges than chutes are employed in embodiments, in order to generate smooth torque, and to make full use of the charge time. The power is proportional to the spin rate, as long as the film thickness can be increased to supply the extra needed current. Unlike a linear actuator (without restoring spring), rotary actuator embodiments can spin much faster than 100 Hz without having the electrostatic force compete with the mechanical accelerations. P33 and S35 isotopes are embodiment candidates when a maximum voltage of 50 kV is imposed. If 300 kV can be achieved without breakdown, then P32 is a selection, however P32 embodiments can require much metal to stop the electrons if only 50 kV is used. Both alpha and gamma emitters were reconsidered. Embodiments achieve at least 50 mW/gm at 1 year life. If operational life were to be restricted to 80 days or less, pure beta emitters are feasible in embodiments. If operational life of 1 year is firm requirement, Sn123 is an available candidate. Note that battery type actuators could provide a potential solution wherein the long life could be achieved through incremental replacement of radioisotope canister. Embodiments may also have manufacturing benefits as well. FIG. 13 depicts a power decay graph 1300 configured in accordance with the present invention. Identified are Sn123 1305; Ca45 1310; P32 1315; and P33 1320. Power=10Log 10((Power/Mass) exp(−t/halflife)) If use of the electrostatic actuator (EA) is after 80 days, embodiments can use Sn123 or Ca45. If use of the EA is before 80 days, embodiments can use P33 or P32. For vertical wall rotor embodiment as in FIGS. 2D and 7B, more torque should be provided for the same source material because the source material is all at a larger radius. For embodiments, there is no advantage to stacking many motors to make double use of the target thickness. Hence, the mass may be double the disk design, when a stacked geometry is used. However, as a stand alone motor, embodiments' power should exceed the source-disk embodiments. In embodiments, current is small at 6 and 12 months. For embodiments, the current is not large enough to support 375 Hz, and get at least 40 kV. Embodiments can have a stroke approximately 20 mm and get 40 kV, but the plate radius would need to be approximately 100 mm. Typically, embodiments employ a minimum stroke distance which allows maximum E field at the running voltage, but in this case the E field at 100 Hz never gets above 2.5e7 V/m, so embodiments increase stroke to improve voltage. Embodiments increase the full stroke from 3 to 6 mm to recover voltage. At higher frequencies, embodiments employ advanced techniques to keep the voltage high: partial discharge, battery, or storage capacitor. Partial discharge methods of operation are beneficial in embodiments in several ways and can result in an overall increase in actuator net power delivered. FIG. 14 depicts maximum beta energy 1400 of source disk by radioisotope in accordance with the present invention. FIG. 15 depicts a partial discharge embodiment with low current analysis figures 1500 configured in accordance with the present invention. It comprises one chute and four source vanes. Views depict perspective FIG. 15A, side FIG. 15B, and top FIG. 15C. The single motor embodiment uses a 3 cm diameter. This rotary embodiment is optimized for low currents at long timelines. Only one chute is used because the current can only charge one chute per revolution. Four source vanes are used to reduce dead time when a vane is being discharged. This embodiment can use either full or partial discharge. Partial discharge is employed at longer timelines and lower currents if frequencies above 50 Hz are desired. The source vane is attracted into the chute, because charges are attracted to higher E fields. When at the throat of the chute, the source vane can be either fully or partially discharged. In embodiments, the partial discharge mechanism could take the form of an LRC circuit with an appropriate time constant. FIG. 16 depicts a one chute, four source vanes, stacked rotary motor 1600 configured in accordance with one embodiment of the present invention. Views depict perspective FIG. 16A, y-axis FIG. 16B, x-axis FIG. 16C, and top FIG. 16D with shared target plates 1605. This figure demonstrates how the rotary embodiment can be stacked to increase the power and increase the P/M. Regarding increasing P/M, each metal target plate is doubly used when the motors are stacked on top of each other. Hence, stacking reduces the target metal mass by half (increasing the P/M by up to a factor 2). This advantage is also present for stacked parallel plates. The smooth symmetry of the chutes is conducive to stacking, because both the top and bottom of the target metal can be used as a chute. Regarding increasing the power, the power is proportional to the number of rotary motors. FIG. 17 depicts Sn123 partial discharge with low current graphs 1700 configured in accordance with one embodiment of the present invention. Curves of FIG. 17A depict P/Ms versus frequency in Hz for 1 year 70nC, 1 year no charge, ½ year 70nC, and ½ year no charge. Curves of FIG. 17B depict P/Ms versus initial charge per vane in nC for 1 year 300 Hz, 1 year 100 Hz, ½ year 300 Hz, and ½ year 100 Hz. Parameters comprise: element=Sn123; Life=1.00 (Yr); Smax=6.0 (mm); 5 min=0.4 (mm); dGap=0.8 (mm); Radius=15.0(mm); Nchute=1.0; Nvane=4.0; Frequency=100.0 (Hz); Emitter Thickness=0.50 (mm); Collector Thickness=0.60 (mm); Structure Thickness=0.40 (mm); Initial Charge=6e-008 (C/vane); Max Voltage =150639.1 (V); Max E field=31.2 (MV/m); Min Current Factor=0.329 (normalized); Power=316.5 (mW); Emitter Mass=1.806 (gm); Collector Mass=3.800 (gm); Structure Mass=1.711 (gm); Total Mass=7.317 (gm); and Specific Power=43.259 (mW/gm). These curves show, for Sn123 embodiments, the advantage of partial discharge when embodiments have low current. For embodiments, when running at very low frequencies, the partial discharge technique is not necessary. The current is allowed to fully charge the EA. In embodiments, if the frequency is too low, the current has too much time to charge the EA and voltage saturation is reached. Hence, no additional power is generated per cycle using extremely low frequencies, and P/M goes to zero as frequency goes to zero. For reference, for Sn123, at 1 year, the potential power output (average decay energynumber decays per kg) is 2100 mW/gm, and at ½ year the potential power output is 5600 mW/gm. Hence, these embodiments get 3.1% efficiency at either of these times. To run at greater than 100 Hz, embodiments use partial discharge (about 10% discharge). Using partial discharge, the P/M is nearly independent of frequency because P/cycle˜VdV˜VIdt, and P=F(P/cycle)˜(1/dt)VIdt=VI. Without partial discharge, with V<Vopt, then V˜It, so P/cycle˜V^2˜I^2dt^2. Then P˜F(P/cycle)˜(1/dt)(I^2 dt^2)˜I^2/dt. One advantage of using partial discharge is the ability to run at higher frequencies with little penalty. No mechanical conversion loss is generated due to a transmission. For partial discharge, as shown in FIG. 17B, an initial charge of 60 nano Coulombs for this embodiment geometry will optimize the P/M, independent of frequency. FIG. 18 depicts rotor partial discharge with low current 1800 configured in accordance with one embodiment of the present invention. Rotor embodiments can have an advantage of partial discharge when they have low current as depicted in the figure and below equations. τ = + ⅆ C ⅆ θ ⁢ ( Q 2 C ) 2 - ⅆ C ⅆ θ ⁢ ( Q 1 C ) 2 = + ⅆ C ⅆ θ ⁢ ( Q 1 + Δ ⁢ ⁢ Q C ) 2 - ⅆ C ⅆ θ ⁢ ( Q 1 C ) 2 Eq . ⁢ 3 τ = + ⅆ C ⅆ θ ⁢ ( Q 1 ⁢ Δ ⁢ ⁢ Q C ) + ⅆ C ⅆ θ ⁢ ( Δ ⁢ ⁢ Q C ) 2 ≈ + ⅆ C ⅆ θ ⁢ ( Q 1 ⁢ Δ ⁢ ⁢ Q C ) Eq . ⁢ 4 FIG. 18 depicts metal target plates 1805 and location of partial discharge 1810. Capacitor 1815 depicts partial discharge target re-charging storage capacitor. The partial discharge term in the equation above is (Q1ΔQ/C). The full discharge term (much less than partial discharge term) is (ΔQ/C)2. This figure explains the advantage of partial discharge, in order to improve P/M. Consider plates 1 and 2. Each is in the same symmetry location inside the symmetric chute, hence each experiences the same magnitude of the slope of the capacitance dC/dth. As the source vane plate 1 moves to the location of source vane plate 2, the charge and voltage is increased 10% to 20%, due to the current. Hence, there is more positive torque at the plate 2 location, compared to the negative torque at the plate 1 location. For partial discharge, the QΔQ term is much larger than the ΔQ2 term, because ΔQ is small compared to Q. FIG. 19 depicts graphs 1900 for rotary actuator, Sn123, ½ year, 375 Hz, partial discharge, configured in accordance with one embodiment of the present invention. Graphs depict an E field slightly larger at end of cycle, which yields net torque in one direction FIG. 19A; voltage which should oscillate around the level of maximum VI product FIG. 19B; current FIG. 19C; and torque slightly larger at end of cycle FIG. 19D. Parameters comprise: element=Sn123; Life=0.50 (Yr); Smax=6.0 (mm); 5 min=0.4 (mm); dGap=0.8 (mm); Radius=15.0 (mm); Nchute=1.0; Nvane=4.0; Frequency=375.0 (Hz); Emitter Thickness=0.50 (mm); Collector Thickness=0.60 (mm); Structure Thickness=0.40 (mm); Initial Charge=7e-008 (C/vane); Max Voltage=168092.2 (V); Max E field=33.8 (MV/m); Min Current Factor=0.298 (normalized); Power=825.2 (mW); Emitter Mass=1.806 (gm); Collector Mass=3.800 (gm); Structure Mass=1.711 (gm); Total Mass=7.317 (gm); and Specific Power=112.783 (mW/gm). Current is limited in the middle of the cycle, at largest gap, when the voltage is highest. The current is strong enough to increase the charge by about 10% during the cycle, above the partial discharge residual charge. The E field is proportional to the charge. The torque increases as the charge squared, and increases by 20% in this example due to the 10% increase in charge. FIG. 20 depicts additional graphs 2000 for rotary actuator, Sn123, ½ year, 375 Hz, partial discharge configured in accordance with one embodiment of the present invention. Graphs depict capacitance per source vane FIG. 20A; gap narrows at the partial discharge locations FIG. 20B; and charge increases 10% over the cycle time, over the initial charge FIG. 20C. The capacitance increases when the source vane is in the throat of the chute, where the gap is the smallest. A symmetric capacitance may appear curious. In embodiments, for maximum power, the capacitance would become very large, with no slope, immediately after the source vane exits the throat of the chute. However, this case is improbable. The source vane, similar to the parallel plate example, always has capacitance to the plate it was just using to pull itself forward. In computer FEM simulations, the slope of the capacitance away from the chute, no matter the exact geometry, was similar to the slope of the capacitance into the chute. Hence, to avoid sharp edges and to keep the analysis somewhat analytical, a smooth taper in and out of the chute was chosen for embodiments, again similar to the parallel plate behavior. FIG. 21 depicts a graph 2100 for rotary actuator power with partial discharge configured in accordance with one embodiment of the present invention. Curves are instantaneous power 2105 and average power 2110. Instantaneous power is much larger than average power (difference between areas of both sides), using partial discharge. This results in an inefficient electrical conversion. Still, embodiments get more net power using partial discharge. Beta energy is going into crossing the gap, instead of heating the target metal. Line 2105 shows the instantaneous power using partial discharge. When the source vane is exiting the chute with the residual charge, there is almost as much work being performed as when the source vane is driving with the torque at the second half of the cycle. The difference between these powers in the negative and positive directions is the net power derived from the EA. As is evident, power is wasted to generate the small average power. However, notably, in embodiments, if partial discharge were not used, this instantaneous energy shown above would exist but as heat in the target metal instead of being used to cross the gap. FIG. 22 depicts finite element model (FEM) results 2200 for capacitance vs. rotation, full discharge, all vanes present configured in accordance with one embodiment of the present invention. Graphs are capacitance vs. rotation angle FIG. 22A and capacitance vs. time period FIG. 22B. Depicted are entrance into chute 2215, leaving chute 2220, and into chute 2225 for analytic model FIG. 22B. The graphs display the capacitance versus rotation angle, for FEM results and for the numerical analysis, for a 1 cm diameter rotary motor embodiment with full discharge. The analytic model 2210 gives similar behavior to the FEM 2205 results. The large roll-off in the capacitance around −10 degrees is due to leaving the chute. For embodiments, the voltage should be low during this part of the cycle, so the reverse torque will be much smaller than the attractive torque into the chute later in the cycle. FIG. 23 depicts capacitance vs. rotation graph 2300, partial discharge, for a large gap, all vanes present, configured in accordance with one embodiment of the present invention. Curve segments include going into chute 2305 and leaving chute 2310. This embodiment geometry has a large gap in the throat of the chute, because partial discharge is assumed, for the 1 cm diameter motor. The slopes leaving and entering the chute are similar; hence the partial discharge technique will work. Note that the geometry is much different leaving the chute, compared to entering the chute, but the capacitances are the similar. The sharp edges in this embodiment may experience arcing. In contrast, the 3 cm diameter rotary embodiment for low current uses a smooth ramp in the target metal into and out of the chute. FIG. 24 depicts graphs 2400 for a low current embodiment configured in accordance with the present invention. Graphs are capacitance vs. rotation FIG. 24A and capacitance vs. time period FIG. 24B. This figure compares analytic capacitance model 2410 with FEM 2405, for the 3 cm diameter rotary motor. Into chute segments are designated 2415 for FIGS. 24A and 2420 for into chute for FIG. 24B. The graphs demonstrate that the magnitudes and slopes of the torque used in the analytic model are reasonable. FIG. 25 are graphs 2500 of Paschen curves. Depicted are breakdown levels with EA 2505, high pressure region 2510, and unfavorable pressures approximately 1/1000 atmosphere 2515. Use of vacuum for embodiments is derived from operation at high voltages. Breakdown voltage is traditionally represented as a function of pressure by the Paschen curves. As illustrated, air at standard pressure and temperature falls below the embodiments' breakdown voltage level. A vacuum condition meets the necessary values. Vacuum is employed in embodiments to achieve the highest power densities. Gas or dielectrics between plates will ionize due to the beta current, probably causing leakage current, and stopping some of the beta electrons from crossing. From the Paschen curve, for embodiments, vacuum needs to be approximately below 1 mm Hg, or about 0.001 atmospheres. Some embodiments may be worse off with a crude vacuum (at minimum on Paschen curve) than with 1 atmosphere. This is because the time between collisions is long enough that the kinetic energy of the ionized particles becomes large enough to ionize other particles. At highest vacuums, ionized particles are “ballistic”; they do not collide with other particles but only collide with the opposite metal plate. From independent collision analysis (estimate the probability of not inducing an ionizing collision as the beta electron crosses the gas), the vacuum requirement for embodiments is estimated to be on the order of approximately 0.001 atmospheres or better to avoid ionizing collisions. FIG. 26 depicts a table 2600 of ionization and ranges of alpha and beta particles. 50 keV beta electrons travel 3.8 mm in air, indicating that there is significant ionization of the air. Faster electrons ionize less because there is less contact time. The significant ionization occurring in air due to the beta electron is an indication that the ionized/conductive gas will not allow a charge buildup on the opposite plates. Predicated upon beta electron emission and capture characteristics, operation of parallel plate actuator embodiments employs high voltage, in excess of 100 kV, including a vacuum envelope. Operation of actuator embodiments at high voltage is feasible. Size is compatible with MAV/NAV operation. Motion transfer with a vacuum envelope may include a magnetic clutch or direct link to actuator bellows for motion transfer. Getters may be used in embodiments to handle material outgassing in order to preserve vacuum qualities and operate at high voltage. Operational voltages for embodiments fall into the category of high voltage and may employ techniques in order to prevent arching. Embodiments can be fabricated using the 3-D Micro Electromagnetic Radio Frequency Systems (3D-MERFS) process. Motion transfer outside the vacuum envelope can be through cam and follower for rotary motor embodiments. In neither case is the operation at high voltage impeded through attachment of the connection mechanism. FIG. 27 depicts graphs 2700 illustrating specific power of beta and gamma emission sources for actuators. Graphs are beta and gamma sources power: specific power (dB) to 1 year FIG. 27A and specific power (dB) vs. days, no gamma FIG. 27B. Elements include Sn123 2705, Ca45 2710, S35 2720, Sr89 2725, and Cd115 2730. Note, however, that although use of a pure beta emitter has been selected in some embodiments for safety purposes, additional embodiments are envisioned. Other active sources can easily be accommodated provided safety considerations are accommodated. Thus, embodiments can use gamma or alpha emission sources as indicated by Sn123, Y91, Sr89, and Cd115 elements. The beta emission process comprises a source material emitting electrons which are then captured by a target material. During the release and capture process, electrical work is performed which in turn is transferred into mechanical work in the form of rotation of the rotor. FIG. 28 is a graph 2800 illustrating power decay of beta sources for actuators configured in accordance with one embodiment of the present invention. Power/mass versus days is depicted for P33 205, P32 210, Ca45 215, and S35 220. Power=10Log 10((power/mass) exp(−t/half life)). All beta radioisotopes are close to equal at 80 days. If they are used EA after 80 days, embodiments employ Ca45. If embodiments use the electrostatic actuator (EA) before 80 days, embodiments use P33 or P32. Therefore, depending on operational life time requirements, a specific beta emitter is chosen. Higher specific powers are achievable with sources which decay quicker, such as P32 or P33. However, longer operational lifetimes are possible with materials such as Ca45 and S35 if the lower specific power is sufficient in terms of achieved MAV/NAV performance. FIG. 29 depicts 2900 an element current chart FIG. 29A and radioactive decay table FIG. 29B. In the chart, the current at time zero (blue bars) is much larger than the current at ½ year and 1 year. The decay process table FIG. 29B depicts alpha decay: current low 2905, P neutron decay: no charge transfer 2910, (electron) emitted, embodiment choice 2915, and monoenergetic, intensities not 100%, gamma decay likely 2920. Alpha decay has high energy, but the penetration depth out of the source film or into the target is very small (a few micro-meters, even with MeV alpha energies). Hence, the source film is very limited in thickness, which imposes a low limit on the current. One possible advantage of using Helium nuclei as charge carriers is that the secondary electron emission, which counteracts the desired current, may be less because a He-electron collision transfers much less energy to the electron than an electron-electron collision. Beta-decay is a candidate decay mode of electrostatic actuator embodiments. The penetration depth is relatively deep in a metal, and the source film is much less limited in thickness/current than alpha decay. Also, beta particles (electrons) will not travel very far in air or through skin, so the particles are safe, from a health standpoint. Internal Conversion is a candidate for the charge mechanism. An excited nucleus decays by transferring its energy to an inner atomic electron, which is ejected with a monoenergetic energy, unlike typical beta decay. However, the intensities are not 100% (probably less than 50%) and the extra intensity goes toward gamma emission. For embodiments, specific power was calculated using: Sp=<E_Peak>fc/m/T Where: <E_Peak>=Peak Beta Kinetic Energy m=Atomic mass of the source T=Half life of source material, and fc=Collection efficiency factor. This takes into account approximate energy spectrum and fixed collection voltage fc=0.073, held constant. Ca45 is a contender for embodiments because the X-ray energy is relatively low, at 12 keV, and the X-rays are only generated 3 out of every million decays. Secondary X-ray emission from the target metal plates and from surrounding support structures, due to collisions with the beta particles, might yield X-rays of comparable energies and higher flux than Ca45. Hence, for embodiments, the target metal and additional structural metals need to be defined, with regard to allowable X-ray creation, before the low flux Ca45 X-ray is deemed a hazard or acceptable. Alpha particles can have large energy (1 MeV) but still the penetration depth out of the source material is very limited. Hence, the film thickness is very limited and the current is low. This low current results in low energy density. Another consideration is beta energy. The power of EA embodiments is largely driven by the maximum voltage (force ˜V2), and voltages less than 10 keV may not generate significant electrostatic force. FIG. 30 depicts energy graphs 3000 configured in accordance with one embodiment of the present invention. Graph FIG. 30A includes probability density showing mean energy 3005. Normalized average collection energy FIG. 30B shows embodiment collection point 3010. Beta electrons exhibit a spectrum of kinetic energies between zero and a peak energy E. The probability density p(E) of this distribution as a function of the kinetic energy E, may be roughly approximated byp(E)=3(1−E/EP)2 Eq. 5 This distribution is shown in the figure. It is heavily weighted towards the lower-energy end of the spectrum. From this distribution, it may be determined that the average kinetic energy of a beta electron is approximately EP/4. If beta electrons are collected at the voltage VC, 0<=eVC<=EP, then the average collected energy per electron, EC(VC), is given byEC(VC)=eVc∫p(E)dE=eVC(1−eVC/EP)3 Eq. 6 from Ep to eVC where e is the electron charge. This collected energy, normalized to EP, is a function of VC, normalized to EP/e. It has a maximum value ofmax{EC}=27EP/256 Eq. 7 at eVC=EP/4, the mean kinetic energy of the beta electrons. Thus, the optimal collection voltage is EP/(4e). Assuming that there are N beta electron emitters per unit volume at their birth, then the beta electron generation density G at the birth of the emitters is given byG=Nln(2)/T Eq. 8 where T is the half life of the emitter. If all beta electrons are collected at the optimal voltage, then the maximum power density Pc that can be obtained by collecting the electrons is given byPC=max{EC}G=0.073EPN/T Eq. 9 The power generation density PC divided by the mass density of the beta electron emitter is then the specific power PS of that emitter. The specific power is a useful metric for comparing different beta electron emitters. Since the mass density of the beta electron emitter is mN, where m is the atomic mass of the emitter,PS=0.073EP/T/m Eq. 10 For some embodiments, current multiplication methods may be applied. Secondary electron emission from metal plates is possible, due to impact with beta electron. Solar cell analogy: a radioisotope can excite electron-hole pairs and induce a current source in the PIN junction. High voltages are not necessary, and the kinetic energy of the beta electron is used to ionize thousands of atoms and create a current. The voltage is determined by placing many PIN junctions in series. Coil transformer, from high impedance/low current to low impedance/high current analogy: have a radioisotope capacitor in a vacuum, and charge to the maximum voltage possible as determined by the isotope. Then discharge into a coil transformer at a very fast rate. A high current/low voltage conversion might be possible in embodiments, which could drive a low voltage electrostatic actuator or a motor. FIG. 31 depicts a table FIG. 31A and graphs FIG. 31B and FIG. 31C of target radiation characteristics 3100. Secondary photon and X-ray spectrum of targets is considered. The lighter elements have the lower energy K-shell X-rays. Be has a 0.111 keV X-ray, which is very minor and would pass right through a human with very little probability of absorption. Note from the X-ray spectra shown on the right, that the resonant K-shell energies are clearly distinguished. This indicates that the background “breaking radiation” is very low at these 1 keV or larger energies. FIG. 32 depicts a further table FIG. 32A and graph FIG. 32B of target radiation characteristics 3200. In embodiments, it is advantageous to have the incident beta electrons pass through the surface of the target metal as quickly as possible, in order to have less likelihood of transferring energy to an electron near the surface. Hence, the beta electron should still have a few keV or more of energy remaining when it collides with the surface of the target metal. (In a “soft” electrostatic collision, less momentum is transferred when the collision is shorter, due to less impulse, i.e., same repulsive force between two electrons but less time for them to interact with each other.) According to the chart, if the incident beta electron strikes the target metal with approximately 200-500 eV of kinetic energy, then there will be on the order of 1 secondary electron per incident beta electron. If the incident beta electron strikes the target with over 2 keV or so, then there will be much less than 1 secondary electron per incident beta electron. The tables of secondary electrons provide information on the energy EII when the ratio of the secondary to incident electron is 1, which typically occurs between 300-2000 eV. Hence uncertainty when the ratio goes well below 1 when the incident beta electron has many keV of kinetic energy. Rougher surfaces have less secondary electrons, for example, compare graphite and soot. Surface roughness creates more surface area to re-capture any ejected secondary electrons. However, this surface roughness may encourage corona currents or gas breakdown, due to charge and E field buildup at sharp points. Copper has more secondary electron emission than Aluminum at 1.5 keV, and, presumably, at all higher energies, due to higher electron density at the surface. Embodiments employ a surface film of a light metallic element. Another source of secondary electrons, besides collisions on the surface, is Auger electrons. When the incident beta electron loses enough energy that its velocity is relatively slow but it still has a few keV of energy, then the probability that the beta electron will excite a resonant energy level within the atoms goes up. The beta electron collides with an inner electron and ejects it, creating a vacant inner shell. A valence electron then drops to the lower energy level to fill the vacant inner shell, and, in doing so, transfers this resonant energy to another electron, which then might escape from the target metal. Hence, in embodiments, we want the incident beta electrons to penetrate well beyond the surface of the target metal before they lose enough energy to excite these Auger electrons. This is another reason to allow the beta electrons to retain more than a few keV of kinetic energy before colliding with the target metal, and is another reason to use light-element target metals, which have low resonance K-shell energies. Hence, the electron will penetrate deeper before exciting the Auger electrons. Auger electrons are used as a tool to identify or characterize elements within materials, so they can be ejected from the surface of the material. FIG. 33 depicts isotope summary tables 3300. Table FIG. 33A parameters include Collection Voltage=˜0.25; Collection Current=˜0.3; Hence VI Power ˜0.250.3=0.075; Energy remaining after escaping from source film=0.7; Best Case fraction of total mass is source mass=0.5. The table presents a list of beta electron emitters, some of their properties, and realizable power outputs using an EA. These are the emitters that: (1) have half lives in excess of 10 days; (2) are not alpha emitters or strong gamma emitters and (3) do not decay into strong alpha or gamma emitters. As a result, Sr90, Ru106 and T1204 are omitted from the table because they all decay to products that are strong gamma emitters. The table however, does include Ca45, which is itself a gamma emitter with the small probability of less than once per 108 events and Sn123 which is a gamma emitter with excellent half life characteristics and limited decay products. PS represents the greatest electrical power per mass that can be extracted from a beta electron emitter by capturing the electrons. In practice, however, this specific power can not be achieved by an actuator for several reasons. First, PS as derived does not include the mass of the collector, the core actuator or its package. Second, a working actuator can not collect all, or perhaps even many, electrons at the optimal collection voltage. This voltage might also, for example, exceed breakdown limitations. Third, no actuator can achieve 100% efficiency. Finally, secondary electron emission at the collector, or any gas within the actuator that is ionized by collisions with the beta electrons, might shunt collected power. Therefore, PS as given, may be an optimistic case. Values for PS are given in Table FIG. 33A above. From the table it is apparent that, for embodiments, four beta electron emitters can meet the specific power requirements of 0.1 W/gm when fabricated into an actuator. They are P32, P33, S35 and Ca45. However, as mentioned above, Ca45 is also a gamma emitter, although at a very low level. Further, P32, P33 and S35 meet the specifications with a sufficient margin to permit the likely fabrication of a practical actuator for <1 year operation. For embodiments, Sn123 meets both the minimum acceptable power density and operational life considerations. Regarding conversion efficiency, the optimal collection voltage is about ¼ the peak voltage, and this allows approximately 0.3 of the beta electrons to cross the gap. This ¼ peak voltage yields the highest VI power product. If the voltage is higher, then the current drops dramatically. The source film for embodiments needs to be about half the penetration depth, in order to yield the maximum current. Hence, some energy is lost as the beta electron escapes from the source film. The energy lost from half the penetration depth was assumed to be 30%, not 50%, because most energy is lost when the electrons are slower and have larger interaction times. The best case mass fraction for embodiments is 0.5. Half the mass is source material, and half the mass is target material, which is not radio-active. The product of these four terms—voltage fraction, current fraction, escape energy, and best case mass ratio—is 0.26. For Table FIG. 33B, Average VI power to Max VI instantaneous power=0.36 Average Mechanical Power to Average VI power=0.78 and Best Case fraction of total mass is source mass=0.5. FIG. 34 depicts graphs of P32 rotary actuator power 3400 configured in accordance with one embodiment of the present invention. Graph FIG. 34A depicts 20 vanes for rotary embodiment seven, FIG. 2C. Graph FIG. 34B depicts 4 vanes for rotary embodiment seven, FIG. 2C. Parameters comprise: a frequency of 100 Hz, Voltage of 55 kV, E max of 2.5e7 V/m, and Gap of 2 mm. Results are 5 Watts at 55 kV using 20 vanes and 22 Watts at 330 kV using 4 vanes. FIG. 35 depicts a graph 3500 of isotope energy power to mass ratio (P/M) versus number of vanes configured in accordance with one embodiment of the present invention. Parameters comprise: a frequency of 100 Hz, Voltage of 50 kV, E max of 2.5e7 V/m, and Gap of 2 mm. The graph shows maximum mechanical power/mass (P/M) in mW/gram for P32, P33, S35, and Ca45 as a function of the number of device vanes. Maximum mechanical power is assumed to be ¼ the maximum VI power. Increase film thickness as the number of vanes increases. Can not increase thickness beyond the penetration depth, so, in embodiments, some elements may be limited in the number of vanes. Embodiments would get double the Power/Mass if rotate at 200 Hz. Both P33 and P32 can tolerate the extra film thickness. It represents an advantage of a rotary actuator. The power density is only limited by the current that can be generated, which is determined by the usable thickness of the source radioisotope film and by the half-life. If the current can be increased substantially, such as with P32 and P33, then rotary actuator embodiments can be made to spin at greater rates, say 500 to 1000 Hz, for example. Embodiment power densities will then be 5 to 10 times larger than those shown in the graph. The graph also indicates an advantage of only using a beta decay energy that is just above the voltage that will be obtained in embodiments. Hence, the metal target thickness will be at a minimum for the amount of energy it is stopping. For example, if P32 is used without consideration of the stopping metal thickness, and if the voltage is limited to 50 kV, then significant metal, hence added weight, is included in embodiments, without benefit. Since the rotary motor can run at 50 kV, embodiments are candidates for materials such as P33, for example, which is a choice from the current and energy perspective in embodiments. The rotary actuator power can be restricted in embodiments if the maximum current from the isotope is weak and the rotary actuator can not spin very fast. Hence, the rotary actuator is very good during the initial timeline of the decay (within one half-life). Following is a detailed comparison between the elements, using S35 as the baseline. S35 ISOTOPE For embodiments, the maximum voltage using Sulfur 35 is probably limited to approximately 40 kV, which means that the target metal can be the thinnest for the isotopes examined here. This is approximately ¼ of the maximum beta energy, and is the optimal collection voltage. For embodiments, Sulfur 35 can have a maximum source thickness of approximately 0.2 mm, because the beta electrons need to be able to escape from the source material, and the penetration depth of the beta electrons, at the maximum 167 keV, is only a little larger than 0.2 mm. Using an 87 day half life, this 0.2 mm source thickness is enough to charge 4 times during a 100 Hz cycle; hence there can be 4 chutes. For embodiments, there also needs to be excess kinetic energy on the beta electron to be able to cross the voltage gap. Charge time should be relatively insensitive to the source wedge width, because the capacitance increases proportional to the width. Hence, if we have a limited current, due to restricted film thickness (due to 167 keV beta energy) and longer half-life (half life of S35 is 87 days), the charge time will be longer and embodiments are limited to only a few chutes or parallel plates. For S35, embodiments are limited to 4 chutes based on the limited current. Using 4 chutes, Power/Mass 300 mW/gm. Ca45 ISOTOPE Calcium 45 has similar behavior to Sulfur 35, except the beta energy is 258 keV instead of 167 keV, and the half-life is 162 days instead of 87 days. The film thickness can be 50% thicker due to the large beta energy, but the total current will be around 0.75 the S35 current due to the factor of two longer half-life. The current is enough to charge the capacitor 3 times during a 100 Hz cycle, and hence there can be 3 chutes. In embodiments, the optimal collection voltage can be 55 kV (or a little larger) instead of 40 kV for S35, which means more power but thicker target metal. Hence, although the longer half life reduces the current by a factor of two, the 30% or so larger voltage compensates for the reduced current. In this figure, 50 kV is assumed, regardless of source, so S35 is assumed to charge to 50 kV instead of 40 kV, and S35 has better Power/Mass than Ca45. However, if we incorporate the 40 kV voltage instead of 50 kV for S35, then Ca45 and S35 would have very close Power/Mass. Using 3 chutes, Power/Mass 170 mW/gm. P33 ISOTOPE P33 has greater differences from Sulfur 35. The beta energy is 249 keV instead of 167 keV, and the half-life is 25 days instead of 87 days. For embodiments, the film thickness can be 50% thicker due to the large beta energy, and the half-life is three times shorter; hence the total current will be around four times larger than the S35 current. The current in embodiments is enough to charge the capacitor 16 times during a 100 Hz cycle, and hence there can be 16 chutes. The optimal collection voltage can be 55 kV (or a little larger) instead of 40 kV for S35, which means more power but thicker target metal. For 16 chute embodiments, by extending the curve, the Power/Mass ˜1000 mW/gm. P32 ISOTOPE P32 has the greatest differences from Sulfur 35. The beta energy is 1700 keV instead of 167 keV, and the half-life is 10 days instead of 87 days. The film thickness in embodiments can be eight times thicker due to the large beta energy, and the half-life is nine times shorter; hence the total current will be around 70 times larger than the S35 current. The current is enough to charge the capacitor 280 times during a 100 Hz cycle, and hence there can be 280 chutes, and the voltage can go up to 400 kV. Other considerations are applied next regarding power. If we just assume that the collection voltage can only be 50 kV based on breakdown, then we only use the large beta energy to create large currents, but we need very thick metal targets to trap the beta electrons (1.2 mm). If 16 chutes are implemented, based on breakdown between source wedges, then, by extending the curve, the Power/Mass 400 mW/gm. For embodiments, P32 would be better than P33 if the actuator can be charged to more than 50 kV. In embodiments, one benefit of the large film thickness is that the source material can be “overstuffed” and an adequate current can be generated for lifetimes 5 or 10 times larger (50 to 100 days) than the element half life of 10 days. FIG. 36 is a flow chart 3600 of a method of a source providing electrons to a target material wherein work is performed configured in accordance with one embodiment of the present invention. The steps comprise transmitting high voltage from a source 3605, high voltage captured by target material 3610, rotational force generated by electrostatic force 3615, and potential discharged 3620. In embodiments, conversion from electrical to mechanical power is accomplished by direct magnetic coupling to the rotor through a vacuum housing. Once outside the housing, a propeller is attached to the far side of the magnetic coupling for direct drive, thus eliminating the losses through a mechanical gear box. The embodiment rotational frequency of the actuator can be adjusted to specifications the propeller such as diameter to optimize propulsion, including efficiency. The foregoing description of the embodiments of the invention has been presented for the purposes of illustration and description. It is not intended to be exhaustive or to limit the invention to the precise form disclosed. Many modifications and variations are possible in light of this disclosure. It is intended that the scope of the invention be limited not by this detailed description, but rather by the claims appended hereto.
	abstract	This fuel assembly for a pressurized water nuclear reactor comprises fuel rods which are arranged at the nodes of a substantially regular network which has a polygonal outer contour, the fuel rods containing uranium which is enriched in isotope 235 and not containing any plutonium before the assembly is used in a reactor. The rods are distributed in at least a first central group which is constituted by fuel rods which have a first level of nuclear reactivity, and an outer peripheral layer of fuel rods which have a level/levels of nuclear reactivity which is/are strictly less than the first level of reactivity.
	claims	1. A detector apparatus for scanning of and obtaining radiation data from an object comprising:a radiation source;a radiation detector system spaced therefrom to define a scanning zone and to collect in use a dataset of information about radiation incident at a detector after interaction with the object in the scanning zone and adapted to resolve such collected information spatially in two dimensions across a scanning area and spectroscopically across a plurality of frequency bands in the spectrum of the source;wherein the detector system is adapted to resolve such collected information spectroscopically in that the detector exhibits a spectroscopically variable response across at least a part of the spectrum of the source; andwherein the detector system is adapted to resolve such collected information spatially in that it comprises:a rastering module configured to divide the scanning area into a plurality of pixels in each of two dimensions at a plurality of resolutions; anda detector control means to move the detector across the scanning area to scan such pixels successively and thereby collect a dataset for each pixel. 2. An apparatus in accordance with claim 1 wherein the detector system is adapted to resolve collected information spectroscopically across at least three frequency bands. 3. An apparatus in accordance with claim 1 wherein the detector system is configured to detect transmitted radiation in that the apparatus comprises a radiation source and a radiation detector system spaced therefrom to define a scanning zone in a radiation transmission path therebetween and thus collect in use a dataset of information about transmissivity of the object in the scanning zone. 4. An apparatus in accordance with claim 1 wherein the source comprises a source to deliver high-energy radiation, and the detector system is adapted correspondingly to detect radiation in the spectrum of the source. 5. An apparatus in accordance with claim 1 further comprising an image generation apparatus to generate at least a first image from an output of the detector system. 6. An apparatus in accordance with claim 5 further comprising an image display adapted to display at least the first image. 7. An apparatus in accordance with claim 5 wherein the image generation apparatus is adapted to receive intensity data from a plurality of spectroscopically resolved energy bands and display these separate datasets of intensity data as separate images successively or simultaneously. 8. An apparatus in accordance with claim 1 wherein the detector comprises a single pixel detector. 9. An apparatus in accordance with claim 1 wherein the detector system is fabricated from a material inherently capable of exhibiting a spectroscopically variable response across at least part of the spectrum of the source. 10. An apparatus in accordance with claim 9 wherein the detector comprises a semiconductor material selected from cadmium telluride, cadmium zinc telluride (CZT), cadmium manganese telluride (CMT), germanium, lanthanum bromide, thorium bromide. 11. An apparatus in accordance with claim 9 wherein the detector comprises a semiconductor material or materials formed as bulk crystal including a Group II-VI semiconductor material. 12. An apparatus in accordance with claim 11 wherein the detector comprises a semiconductor material selected from cadmium telluride, cadmium zinc telluride (CZT), cadmium manganese telluride (CMT). 13. An apparatus in accordance with claim 1 wherein the rastering module is configured to divide the scanning area into at least a coarse and a fine resolution, and is further configured to enable selection between the resolutions by a user through a suitable user input interface. 14. An apparatus in accordance with claim 1 wherein the radiation is at least of one x-rays, gamma rays, and subatomic particle radiation. 15. An apparatus in accordance with claim 1 wherein the radiation source is adapted to emit a radiation beam over a wide area, and wherein the detector has a reduced detection area that is configured to cover and differentiate across only a small part of the overall area to be scanned at any given time. 16. A method of obtaining radiation interaction data from an object comprising the steps of:providing a radiation source and a radiation detector system spaced therefrom to define a scanning zone therebetween, wherein the detector system is adapted to resolve collected information spectroscopically in that it comprises a detector that exhibits a spectroscopically variable response across at least a part of the spectrum of the source;defining a scanning area for collection of radiation incident at the detector;dividing the scanning area into a plurality of pixels in each of two dimensions at a plurality of resolutions;moving the detector across the scanning area to scan such pixels successively and thereby collect a dataset for each pixel of information about radiation incident at the detector after interaction with the object in the scanning zone; andresolving each such dataset spectroscopically across a plurality of frequency bands within the spectrum of the source. 17. A method in accordance with claim 16 wherein the detector is adapted to differentiate information spatially to a resolution of fewer than the said number of pixels and/or is dimensioned to less than the desired scanning area of the dataset such that the spatial resolution in the collected dataset is attributable at least in part to the step of moving the detector across the scanning area to scan pixels successively in each of two dimensions at a plurality of resolutions. 18. A method in accordance with claim 17 wherein the detector is a single pixel detector, and the spatially resolved dataset is assembled by scanning each pixel individually. 19. A method in accordance with claim 16 comprising the steps of providing a radiation source and a radiation detector system spaced therefrom to define a scanning zone in a radiation transmission path therebetween, and thereby collecting a dataset of information about transmissivity of the object in the scanning zone. 20. A method in accordance with claim 16 comprising a preliminary step of scanning/imaging the object using any suitable apparatus to identify areas of interest in the object and wherein steps of claim 16 are subsequently performed to further investigate such areas of interest. 21. A method in accordance with claim 16 further comprising the step of generating an image of the object in the scanning zone. 22. A method in accordance with claim 21 further comprising the step of displaying such generated image on a suitable display apparatus. 23. A method in accordance with claim 21 wherein each collected image is resolved spectroscopically across a plurality of bands to generate an energy-differentiated composite image or succession of images. 24. A method in accordance with claim 16 performed initially at a first coarser resolution to identify areas of an object for further investigation and subsequently at a second finer resolution collecting data only from those areas. 25. A method in accordance with claim 16 wherein the radiation source is an x-ray or a gamma ray source. 26. A method in accordance with claim 16 wherein the radiation source emits a radiation beam over a wide area, and wherein the detector has a reduced detection area that covers and differentiates across only a small part of the overall area to be scanned at any given time.

Subsets and Splits

No community queries yet

The top public SQL queries from the community will appear here once available.