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	abstract	A coolant injection system for a nuclear power generation system includes the coolant injection system, and method of operation of the coolant injection system. The nuclear power generation system includes a reactor pressure vessel having a reactor core, a pressuriser in fluid communication with the reactor pressure vessel, and the injection system, which comprises a make-up tank having a tank inlet and a tank outlet. The injection system has an operating condition, and a fault response condition, and is configured to switch between these conditions when coolant level in the pressuriser drops below a threshold level. In the operating condition, the tank outlet is isolated from the reactor pressure vessel such that coolant is retained in the make-up tank, and the tank inlet is in fluid communication with the reactor pressure vessel and the pressuriser.
	abstract	A clamp apparatus is provided which is designed to structurally replace a weld that attaches a riser brace assembly to a reactor pressure vessel wall, and a method for repairing the riser brace assembly. The riser brace assembly is designed to support a jet pump in the reactor pressure vessel. The riser brace assembly may include upper and lower riser brace leaves connected to a reactor pressure vessel pad on the wall. The clamp apparatus may include a first clamp component including a central extension portion, and a second clamp component including a slot portion. The central extension and slot portions may be engaged to provide alignment between the first and second clamp components between the upper and lower riser brace leaves of the riser brace assembly.
061887463	summary	BACKGROUND OF THE INVENTION The development of a compact, tunable, hard x-ray source would have profound and wide ranging applications in a number of areas. These areas include x-ray diagnostics, medical imaging, microscopy, nuclear resonance absorption, solid-state physics and material science. Currently, varieties of x-ray generators exist. The most modern devices are generally based on one of three methodologies: laser and discharge plasmas, electron impact sources, and synchrotron. The spectrum of these sources can be divided into two categories: characteristic x-rays and continuum x-rays. The characteristic x-ray sources are dependent on the particular atomic structure of the gas or target material in use. Among all the types of x-ray sources, only synchrotron produces continuum radiation. The main interest in laser-generated plasma is directed towards inertial confined fusion. Recently, they have also gained interest as sources of (V)UV and x-rays. Laser-generated plasmas emit photons in an energy range, which can extend from visible light to hard x-rays. The observed emission spectrum is characteristic of a high-temperature, short-lived, high-density plasma. The sources produce a spectrum of x-rays centered about characteristic lines of the material. In a laser-generated plasma x-ray source, when a high-power pulsed laser is focused on a (solid) target, a plasma is created. After the laser pulse terminates, the plasma cools extremely rapidly due to rapid thermal conduction, electron energy loss to ions, and expansion of the plasma into the surrounding vacuum. Cooling of the electrons at high density leads to fast recombination, quenching of the highly excited states, and a termination of the x-ray emission. The choice of target material controls the intrinsic range of the spectral output determined by the ionization states of the target material. Details of the spectral distribution are highly dependent on the target material (e.g., carbon, aluminum, titanium, copper, zinc, molybdenum, tin, tungsten, and lead) and other parameters (target thickness and source size). Plasma discharge systems have been suggested as sources of high brightness x-ray radiation. Most of these devices (the gas puff J. Pearlman an J. C. Riordan, J. Vac. Sci. Technol. 19, 1190 (1981), plasma focus Y. Kato, et al, Appl. Phys. Lett. 48,686 (1986), and hypocycloidal pinch K. S. Han, et al, Bull. Am. Phys. 31, (1986)) are variations of the Z-pinch geometry. In Z-pinch devices, a high current is produced on the outer edge of a cylindrical volume of gas using a pulsed electrical driver such as a fast capacitor bank. The resulting JxB force accelerates the plasma shell radially inward to form a very high-temperature plasma on-axis which emits characteristic thermal radiation in the soft x-ray region. The conventional electron impact sources use a suitable target material that is bombarded by a high-energy electron beam. These sources produce a broad spectrum of x-rays centered about characteristic lines of the material. Synchrotron radiation is the electromagnetic radiation emitted by electrons moving at relativistic velocities along a curved trajectory with a large radius of curvature, for example, several meters to tens of meters. The energy of the photons ranges from a few electron volts to 10.sup.5 Ev. This corresponds to the binding energy of electrons in atoms, molecules, solids, and biological systems. Thus, synchrotron radiation photons have the right energy to probe the properties of such electrons and of the corresponding chemical bonds to understand their physical and chemical properties. The uses of electron accelerators as sources of synchrotron radiation have grown enormously during the last two decades. Unique features such as tunability and wide x-ray spectrum tend to render the synchrotron irreplaceable for many applications. Presently, third generation synchrotron sources are being pursued that are based on high-energy electron storage rings and bending magnets. A typical electron accelerator can be tuned to emit synchrotron radiation in a very broad range of photon energies, from microwaves to hard-x-rays. Thus, it provides electromagnetic radiation in spectral regions for which no other usable sources exist, e.g., most of the ultraviolet/soft-x-ray range. Furthermore, it is by far the best source of hard-x-rays, even though other sources exist for this range. The system has met most application needs, but fails with respect to physical size and cost. They are inevitably large and expensive devices requiring complex supporting facilities. The current machines are very large and costly with tens to hundreds of millions of dollars. The nature of synchrotron x-ray sources means that they are expensive, remote multi-user facilities, and are therefore not suited for use with a laboratory scale. The alternative x-ray sources, such as electron impact systems, laser and discharge plasmas, cannot match synchrotron in terms of its tunability and continuum x-rays. An object of the invention disclosed is to provide a small compact tunable x-ray source. Another object is to provide a compact tunable x-ray source for laboratory use. For applications where a relatively small sample is practical, the availability of a laboratory-scale source would be very advantageous. Another object is to provide a compact tunable x-ray source for security inspection applications such as more sensitive balcale x-ray inspection systems. SUMMARY OF THE INVENTION A low cost, compact, tunable x-ray source, that is based on an inertial electrostatic confinement (IEC) vessel design, is proposed. The IEC device is described in pending U.S. patent application Ser. No. 08/232,764 for "Inertial-Electrostatic Confinement Particle Generator" and Ser. No. 08/491,127 for "Electrostatic Accelerated Recirculating Fusion Neutron/Proton Source" which are incorporated herein by reference. In the IEC-based x-ray source design, the electron storage ring of the synchrotron is replaced by recirculatory focusing electrons in a sphere that are accelerated by a grid, and the bending magnets are replaced by the electron--electron collisions in the sphere center. This arrangement results in an IEC synchrotron source (IEC-SS), wherein the mechanism for generating tunable x-ray radiation is essentially the same as in the bending magnet synchrotron sources. The IEC-SS operates at a much lower electron energy (<100 keV compared with >200 MeV in a synchrotron) while still giving a same radiated x-ray energy compensated by a bending radius of much smaller scale from electron--electron interactions. In short, electrons are accelerated 10's to 100 kev by the anode grid. Due to spherical (or other) convergence, the energetic electrons scatter in the center of the sphere. The scattering interactions create intense bremsstrahlung x-rays. The emitted x-ray energy is controlled by the grid bias.
	abstract	A hazardous material storage bank includes a wellbore extending into the Earth and including an entry at least proximate a terranean surface, the wellbore including a substantially vertical portion, a transition portion, and a substantially horizontal portion; a storage area coupled to the substantially horizontal portion of the well bore, the storage area within or below a shale formation, the storage area vertically isolated, by the shale formation, from a subterranean zone that includes mobile water; a storage container positioned in the storage area, the storage container sized to fit from the wellbore entry through the substantially vertical, the transition, and the substantially horizontal portions of the wellbore, and into the storage area, the storage container including an inner cavity sized enclose hazardous material; and a seal positioned in the wellbore, the seal isolating the storage portion of the wellbore from the entry of the wellbore.
	description	The present invention relates to a method for inspecting defects of the cladding of a nuclear fuel rod, and more particularly, to a nondestructive method for measuring internal liquid level and detecting leakages of a cladding tube. To be more specific, the method of the invention provides a wave obliquely incident to a tube at a specified incident position for checking whether a liquid exists inside the tube at the incident position by detecting the wave received at the side of the tube opposite to the incident position. Since the fuel rod placed in the reactor core of a light water nuclear reactor is compromised of a Zircaloy cladding tube filled with uranium fuel pellets, once the cladding breaks, the nuclear fuel therein could be released into the reactor as to radioactively pollute the cooling water that, as a consequence, the safety of the whole nuclear power plant is affected and the radioactivity of the material discharged therefrom is increased. In a worse scenario, the operation of the nuclear power plant will have to be shut down for inspection so that causes a severe economic loss. In this regard, a method capable of detecting precisely the defects of a cladding tube can play an important role for preventing the above-mentioned economic loss by avoiding the foregoing radioactive pollution from happening. However, it is not an easy and convenient job for even a trained technician to identify a defect using the current ultrasonic inspection method that is usually the cause of detainment for a scheduled maintenance of nuclear power plant. Therefore, an improve inspection method, which is fast and precise, can save the maintenance cost in millions by the saving of working hours for the maintenance. It is one of the most important safety features for a nuclear power plant to be able to rapidly and precisely detect a fuel rod with defected cladding so as to proceed with the replacement of the defected fuel rod. Refer to FIG. 1A and FIG. 1B, which are schematic illustrations showing a conventional ultrasonic inspection method used by most nuclear power plant for detecting defects of a fuel rod. As seen in FIG. 1A, an ultrasonic signal 111 emitted from an ultrasonic emitter 11 is vertically-incident to the cladding tube 12 of a fuel rod to be inspected that the ultrasonic signal 111 will travel through the internal 121 of the tube and finally to be received by the receiver 13. The magnitude of the ultrasonic signal 111 received by the receiver 13 will increase when there is cooling water accumulated in the internal 121 of the tube caused by a cracking of the cladding 12 enabling the cooling water to enter therefrom. By which, the situation of water accumulated inside a cladding tube can be detected. FIG. 1B shows another method for detecting the situation of water accumulated inside a cladding tube. Similarly, an ultrasonic signal 111 emitted from an ultrasonic emitter 11 is vertically incident to the cladding tube 12 of a fuel rod and further into the internal 121 of the tube. If there exists the situation of water accumulated inside a cladding tube caused by the cracked cladding tube 12, the ultrasonic signal 111 traveling inside the tube will be attenuated and reflected and the reflected signal will be received by the receiver 11′. That is, if an attenuated ultrasonic signal is detected by the receiver 11′, there is surely water accumulated inside the cladding tube. From the above description, the conventional inspection method has the following shortcomings: (1) The reliability of the conventional inspection method is low, since a minute decrease of signal might not be caused by accumulated water resulting from defected cladding. Moreover, it is difficult to identify a variation in magnitude of the ultrasonic signal, since noise will have an effect while evaluating the magnitude of the signal. (2) The convention inspection method can only determine whether there is water accumulated inside the cladding tube, but can not detect the exact position of cracking or defect. (3) The inspection accuracy of the convention inspection method is easily affected by the shape of the object to be inspected. It is more suitable to be used for inspecting a flat object and not for a hollow tube or an object with curved surface. (4) The convention inspection method is substantially a method of two-dimensional measuring, which is not as effective while it is used for inspecting defects of an upright cladding tube.Thus, it is indeed a pressing requirement for improving the conventional inspection method. It is the primary object of the invention to provide a nondestructive method for inspecting defects of the cladding of a nuclear fuel rod, in which an inspection wave is obliquely incident to the cladding tube for checking whether water is accumulated inside the cladding tube by utilizing the refraction phenomenon of the inspection wave while the same traveling between different mediums. It is another object of the invention to provide a nondestructive method for inspecting defects of the cladding of a nuclear fuel rod, which is capable of detecting the liquid level inside the cladding tube by providing an inspection wave obliquely incident to the cladding tube so as to utilize the refraction phenomenon of the inspection wave while the same traveling between different mediums. To achieve the above objects, the nondestructive method for inspecting defects of the cladding of a nuclear fuel rod of the present invention comprises the steps of: (a) providing a tube to be inspected, wherein the tube comprises an outer surface and an internal surface; (b) arranging a wave emitter close to a side of the tube to be inspected for discharging an inspection wave obliquely incident to a first position on the outer surface by a predefined tilt angle and subsequently progressing to come into contact with the internal surface at a second position thereon; (c) arranging a receiver at another side of the tube with respect to the wave emitter for receiving the inspection wave passing through the tube; (d) making an evaluation to determine whether the passing-through inspection wave can only be detected by the receiver while the same is being arranged at a specific position at another side of the tube with respect to the wave emitter; if so, it represents that there is no liquid existed inside the tube under the level indicated by the second position; (e) making an evaluation to determine whether the passing-through inspection wave can be detected by the receiver while the same is being arranged at two different positions both at another side of the tube with respect to the wave emitter; if so, it represents that there is liquid existed inside the tube at the level indicated by the second position. Other and further features, advantages and benefits of the invention will become apparent in the following description taken in conjunction with the following drawings. It is to be understood that the foregoing general description and following detailed description are exemplary and explanatory but are not to be restrictive of the invention. The accompanying drawings are incorporated in and constitute a part of this application and, together with the description, serve to explain the principles of the invention in general terms. Like numerals refer to like parts throughout the disclosure. For your esteemed members of reviewing committee to further understand and recognize the fulfilled functions and structural characteristics of the invention, several preferable embodiments cooperating with detailed description are presented as follows. Please refer to FIG. 2A, which is a schematic illustration of using a nondestructive method of the present invention to inspect the cladding tube 22 of a nuclear fuel rod while there is no liquid accumulated inside the cladding tube 22. As seen in FIG. 2A, a tube 22 to be inspected having an outer surface 221 and an internal surface 222 is provided where a wave emitter 21 is arranged close to a side of the tube 22 for obliquely discharging an inspection wave 211. In a preferred embodiment of the invention, the inspection wave 211 is substantially an ultrasonic wave. After discharging, the inspection wave 211 will be incident to a first position 223 on the outer surface 221 by a predefined tilt angle θ and subsequently progressing to come into contact with the internal surface 222 at a second position 224 thereon. Since the frequency of the ultrasonic wave is ranged between 1 MHz and 25 MHz which is very high and can only be transmitted through a medium. While there is no cracking on the tube and thus no liquid accumulated inside the tube, there will be hardly any medium existed in the tube capable of transmitting the ultrasonic wave used as the inspection wave 211 and thus most of the inspection wave 211 will be reflected while coming into contact with the second position 224 on the internal surface 222 without passing through the same and entering into the tube 22. As a consequence, the inspection wave 211 will progress along the path shown as the dotted line of FIG. 2A and finally out of the tube 22 at a proper position to be received by the receiver 23 arranged at another side of the tube 22 with respect to the wave emitter 21. It is noted that the transmission mechanism defining the path of the inspection wave 211 is very complicated while the same is progressing inside the wall of the tube 22 between the outer surface 221 and the internal surface 222, that the path is highly related to the material of the tube 22 and the incident angle of the inspection wave 211. However, there is still a portion of the inspection wave dissipating out of the tube wall during each reflection as shown by the twist arrows of FIG. 2A. As a matter of fact, the wave received by the receiver 23 is only the signal of the reflected inspection wave with highest intensity that is not exactly the whole original inspection wave 211. Please refer to FIG. 2B, which is a schematic illustration of using a nondestructive method of the present invention to inspect the cladding of a nuclear fuel rod while there is liquid accumulated inside the cladding. Similarly, a tube 22 to be inspected having an outer surface 221 and an internal surface 222 is provided where a wave emitter 21 is arranged close to a side of the tube 22 for obliquely discharging an inspection wave 211. In a preferred embodiment of the invention, the inspection wave 211 is substantially an ultrasonic wave. After discharging, the inspection wave 211 will be incident to a first position 223 on the outer surface 221 by a predefined tilt angle θ and subsequently progressing to come into contact with the internal surface 222 at a second position 224 thereon. Since a liquid 29 is already existed inside the tube 22 that is not the same material as the one of the tube 22, the second position 224 is acted as an interface formed between the liquid and the material of the tube while the inspection wave 211 progresses to the second position 224. In this regard, at the second position 224, a portion of the ultrasonic wave used as the inspection wave 211 is reflected and progresses along a reflection path, i.e. the first path, as the one shown in the dotted line I of FIG. 2B which is finally being transmitted out of the tube 22 at a proper position to be received by the first receiver 23 arranged at another side of the tube 22 with respect to the wave emitter 21. In addition, according to Snell's law of refraction, some other portion of the ultrasonic wave used as the inspection wave 211 is refracted at the second position and enters into the liquid 29 following the path shown in the dotted line II of FIG. 2B which is further refracted by the interface between the liquid and the material of tube 22 so as to be transmitted out of the tube 22 and is received by the second receiver 23′. In retrospect to the above description, if two different passing-through inspection waves 211 respectively can be detected by two receivers 23, 23′ while the same are respectively being arranged at two different positions both at another side of the tube 22 with respect to the wave emitter 21, it represents that there is liquid existed inside the tube. Similarly, there is still a portion of the inspection wave 211 dissipating out of the tube wall during each reflection as shown by the twist arrows of FIG. 2B. As a matter of fact, the wave received by the first receiver 23 is only the signal of the reflected inspection wave with highest intensity that is not exactly the whole original inspection wave 211. In addition, the total intensity of the signal received by the first receiver 23 and the second receiver 23′ is smaller than the original inspection wave 211. Moreover, the signal intensity of the wave received by the first receiver 23 while there is liquid accumulated inside the tube 22 is obviously smaller than that of the wave received by the first receiver 23 while there is no liquid existed in the tube 22, and the signal intensity of wave received by the second receiver 23′ of FIG. 2B is also larger than that of the wave received by the first receiver 23. Moreover, the path of the inspection wave 211 transmitting between the internal surface 222 and outer surface 221 of the tube 22 and the refraction angle defining the refraction of the inspection wave 211 while the same entering the liquid 29 are closely related to the material made of the tube 22, the incident angle θ and the refraction index of the liquid 29. Hence, an elaborate calculation procedure is required for obtaining a precise incident angle for the inspection wave while using the conventional method for inspecting a cladding tube. On the other hand, the method of the present invention is capable of achieving the disclosed inspection object without requiring ascertained parameters. That is, as soon as two obvious signals are detected at a side of the inspected tube with respect to the wave emitter, it can be certain that there is liquid existed in the tube. In another preferred embodiment of the invention, the wave emitter 21 can be connected to a driving mechanism (not shown in the figures), which can drive and position the wave emitter 21 at different locations allowable by the current setting for enabling the wave emitter 21 to discharge inspection wave to the inspected tube by different incident angle. In addition, the first receiver 23 and the second receiver 23′ of FIG. 2B can be integrated into a signal receiving device, which is similarly being connected to a driving mechanism for enabling the receiving device to search the inspection wave actively. Yet, in another embodiment of the invention, the receiving device is stationary, but is arranged with a specified amount of receivers therein, which is also capable of achieving the same effect as the active receiving device. The embodiments described above are all related to a method for inspecting a tube, and more particular, to a method for detecting leakages of a waterproof tube. However, the method of the invention can further be employed by the industry for detecting liquid level of a storage tank storing corrosive liquid or liquid pollutant. In this regard, the wave emitter is driven and positioned by a driving mechanism enabling the emitter to discharge wave to an inspected object at different altitudes while an evaluation is being made for determine whether there is liquid existed at those altitudes according to the amount of signals received by the receiver. Hence, the present invention can be implemented for detecting the level of liquid accumulated in an upright tubular object. Yet, in another preferred embodiment of the invention, light wave can be employed as the inspection wave while the tube 22 is made of a transparent material. The operation of the present embodiment of using light wave as the inspection wave is similar to that of the prior embodiment that is not described further hereinafter. However, the present embodiment of using light wave enables the method of the invention to be adopted massively by the unmanned monitoring apparatus of an automation production system. While the preferred embodiment of the invention has been set forth for the purpose of disclosure, modifications of the disclosed embodiment of the invention as well as other embodiments thereof may occur to those skilled in the art. Accordingly, the appended claims are intended to cover all embodiments which do not depart from the spirit and scope of the invention.
046866949	summary	CROSS-REFERENCE TO RELATED APPLICATIONS This application is related in subject matter to the following co-pending applications: Ser. No. 216,226, filed Dec. 15, 1980, now abandoned, entitled: "Filter Means For An Apparatus For Analyzing Metals By X-Ray Fluorescence" (TN-13-PA-US) and U.S. Pat. No. 4,429,409, entitled "Portable Apparatus For Analyzing Metals By X-Ray Fluorescence" (TN-15-PA-US) and assigned to the same assignee as this application. BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to a device for the identification of metal alloys by the X-ray fluorescence method. 2. Description of the Prior Art Many businesses manufacture products from stock pieces of metal alloys purchased from a manufacturer of such stock. If several different alloys having a similar appearance are being stored and utilized in the daily manufacturing process, a mixed material problem can occur. If the wrong alloy is utilized in the manufacture of a part, it may result in the premature failure of the part during normal use. Such a failure has the potential for serious economic consequences and physical danger. As businesses became aware of the mixed metals problem, they turned to quantitative inspection techniques including X-ray fluorescence. Many types of devices for X-ray flourescence analysis are known. Radiation is emitted by a sealed radioactive source and impinges upon the sample being tested. The radiation initiates the emission of secondary X-radition from the sample. The secondary X-radiation is sensed and the concentration of any element in the sample is determined by the intensity of the characteristic X-rays of the element in the spectrum. Use can be made of special filters which make it possible to eliminate certain spectral lines so that only those that are typical of a given element are permitted to remain. Thus, by using a series of different filters, it is possible to determine the composition and concentration of the constituents of any sample. U.S. Pat. No. 3,992,542 discloses an apparatus for the continuous analysis of samples. The apparatus includes a measuring head having a radioactive source and a counting assembly connected to the radiation detector. A sequential filter transfer unit has a conveyor driven in reciprocating motion between a filter stack and a gap between the source and the detector. A sample transfer unit with inclined parallel slide ramps and a receiving trough fitted with a push plate for passing the sample in front of the source in unitary sequence is controlled by a mechanical control assembly and an electronic assembly for recording signals delivered by the radiation detector after analysis of each sample. However, such a device has the disadvantage of requiring a sample to be brought to the device for analysis. Furthermore, the sample must be in a certain size range in order to be passed in front of the radioactive source. SUMMARY OF THE INVENTION The present invention concerns an apparatus for analyzing a sample of material by the X-ray fluorescence method. The apparatus includes a source of X-ray radiation, a radiation detector, and a plurality of filters adapted to preferentially pass the spectral lines of a predetermined one of a plurality of different elements, positioned in a hand-held probe connected to an electronic unit. The probe includes a hollow, generally cylindrical detector housing enclosing the radiation detector and the filters and a hollow, generally right triangular prism shaped source housing enclosing the radiation source. The source housing tapers from an open base attached to the detector housing to a tip having an aperture formed therein. A pair of radioactive sources are positioned in the housing adjacent and on either side of the tip aperture. Each source is enclosed by a collimator having an exit port directed toward the tip aperture and each collimator is enclosed by a shutter means moveable between a first position blocking radiation and a second position passing radiation from the source to the tip aperture. Radiation from the sources generates X-rays from the sample which X-rays pass through the aperture in the tip and the open base of the source housing and enter the detector housing through an aperture formed therein.
	abstract	In a radiation image conversion panel (10), a radiation conversion layer (2) for converting an incident radiation into light is formed on a substrate (1). The radiation conversion layer (2) has a reflective layer (3), on a side opposite from a light exit surface (2a) for emitting the light, for reflecting the light to the exit surface (2a) side, while the reflective layer (3) has a helical structure comprising helically stacked phosphor crystals. Thus constructed radiation image conversion panel (10) can enhance the reflectance without forming a reflective layer made of a thin metal film or the like and exhibit a reflectance higher than that in the case where the reflective layer is formed by spherical crystal particles.
049845101	summary	The present invention concerns a system for posting articles into a containment which is maintained at sub-atmospheric pressure. In particular the invention concerns a posting system with a port having a sphincter seal for minimising back diffusion through the port during a posting operation. FEATURES AND ASPECTS OF THE INVENTION According to the present invention a system for posting articles into a containment maintained at sub-atmospheric pressure comprises a posting port in the wall of the containment having a sphincter seal and a removable cover or 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. Preferably, the sphincter seal comprises an annular assembly of inner and outer brush seals between which are sandwiched rings of resilient material. The outer annular brushes of the assembly serve to protect the resilient rings during posting operations. The lower annular brushes serve to support the resilient rings and help prevent permanent deformation. Both inner and outer brushes restrict the opening in the containment during posting operations.
	claims	1. A nuclear reactor vessel comprising:an elongated cylindrical body defining an internal cavity containing primary coolant water;a nuclear fuel core disposed in the internal cavity;an elongated cylindrical shroud disposed in the internal cavity, the shroud comprising an inner shell, an outer shell, and a plurality of intermediate shells disposed between the inner and outer shells;a plurality of annular cavities formed between the inner and outer shells by the intermediate shells, the annular cavities each being filled with the primary coolant water;the shroud being comprised of a plurality of vertically-stacked shroud segments each including an annular top closure plate and an annular bottom closure plate, wherein the inner, outer, and intermediate shells of each shroud segment is seal welded to their respective top and bottom closure plates collectively forming a self-supporting shroud segment structure; andeach shroud segment detachably coupled to an adjoining shroud segment. 2. The reactor vessel of claim 1, wherein the annular cavities are fluidly interconnected by a plurality of drain holes formed through the outer shell and intermediate shells, the drain holes in the outer shell spaced longitudinally apart along a length of the outer shell, and the drain holes in the intermediate shells being spaced longitudinally apart along a respective length of each intermediate shell. 3. The reactor vessel of claim 1, wherein the intermediate shells each have a length that is coextensive with respective lengths of the inner and outer shells. 4. The reactor vessel of claim 1, wherein the shroud forms a riser region inside the shroud and a downcomer region between the shroud and the body of the reactor vessel. 5. The reactor vessel of claim 4, wherein the fuel core is disposed inside the shroud. 6. The reactor vessel of claim 4, wherein the shroud has a bottom which is spaced vertically apart from a bottom of the reactor vessel to form a fluid flow path between the downcomer and riser regions. 7. The reactor vessel of claim 1, further comprising a plurality of seismic springs disposed between the shroud and the body of the reactor vessel for lateral restraint. 8. The reactor vessel of claim 7, wherein the seismic springs are arcuately shaped concave leaf springs each comprising a plurality of individual leaves joined together and opposite ends engaging an interior surface of the cylinder body of the reactor vessel. 9. The reactor vessel of claim 1, further comprising a clamp coupling the top closure plate of one shroud segment to the bottom closure plate of an adjoining shroud segment. 10. The reactor vessel of claim 9, wherein the clamp is pivotably mounted to a shroud segment. 11. The reactor vessel of claim 9, further comprising a seismic restraint fixedly attached to the clamp, the seismic restraint configured to engage the body of the reactor vessel. 12. A nuclear reactor vessel comprising:an elongated cylindrical body defining an internal cavity containing primary coolant water;a nuclear fuel core disposed in the internal cavity;an elongated cylindrical shroud disposed in the internal cavity, the shroud comprising a plurality of vertically-stacked shroud segments each including an annular top closure plate, an annular bottom closure plate, an inner shell, an outer shell, and a plurality of radially spaced apart intermediate shells disposed between the inner and outer shells;an outer annular cavity formed between the intermediate shells and the outer shell;an inner annular cavity formed between the intermediate shells and the inner shell;a plurality of intermediate annular cavities formed between the intermediate shells; anda plurality of drain holes formed through the outer shell and each intermediate shell, the drain holes in the outer shell spaced longitudinally apart along a length of the outer shell, and the drain holes in the intermediate shells being spaced longitudinally apart along a respective length of each intermediate shell, the drain holes operable to fluidly interconnect the outer, inner, and intermediate annular cavities;wherein the inner, outer, and intermediate shells of each shroud segment is seal welded to their respective top and bottom closure plates collectively forming a self-supporting shroud segment structure, each shroud segment detachably coupled to an adjoining shroud segment;wherein the outer, inner, and intermediate annular cavities are each filled with the primary coolant water.
	claims	1. A pressurized light water reactor (PLWR) having direct vessel injection (DVI) nozzles for reducing emergency core coolant (ECC) bypass through a broken cold leg comprising:a cylindrical reactor vessel having a central axis;a core support located within the reactor vessel to support a reactor core in the reactor vessel;a first hot leg (4a) and a second hot leg (4b) located on the cylindrical reactor vessel at two diametrically opposed positions along an axis and at a first height positioned above the reactor core;a first cold leg (3a), a second cold leg (3b), a third cold leg (3c) and a fourth cold leg (3d) located on the cylindrical reactor vessel each at said first height above the reactor core;the second cold leg (3b) and the third cold leg (3c) are located on opposite sides of the first hot leg (4a) and spaced from the first hot leg at an angle of 60°, and the first cold leg (3a) and the fourth cold leg (3d) are located on opposite sides of the second hot leg (4b), and spaced from the second hot leg at an angle of 60°;a first DVI nozzle (5a), a second DVI nozzle (5b), a third DVI nozzle (5c) and a fourth DVI nozzle (5d) located on the cylindrical reactor vessel wherein the first DVI nozzle (5a) and the fourth DVI nozzle (5d) are located at a distance above the axis of the first and second hot legs where said distance equals 1.5 times a sum of a diameter of the second hot leg and a diameter of the first DVI nozzle; andthe first DVI nozzle (5a) is located between the second hot leg (4b) and the first cold leg (3a), and closer to the second hot leg than to the first cold leg at a first angle from the second hot leg of 10 degrees up to and including 15 degrees, the fourth DVI nozzle (5d) is located between the second hot leg (4b) and the fourth cold leg (3d), and closer to the second hot leg than to the fourth cold leg at a second angle from the second hot leg of 10 degrees up to and including 15 degrees, the second DVI nozzle (5b) is located between the first hot leg (4a) and the second cold leg (3b), and closer to the first hot leg than to the second cold leg at a third angle from the second hot leg of 10 degrees up to and including 15 degrees, the third DVI nozzle (5c) is located between the first hot leg (4a) and the third cold leg (3c), and closer to the first hot leg than to the third cold leg at a fourth angle from the first hot leg of 10 degrees up to and including 15 degrees. 2. The pressurized light water reactor (PLWR) according to claim 1, wherein each of the first angle, the second angle, the third angle and the fourth angle are 15°.
	description	This application claims the benefit of U.S. Provisional Application No. 61/792,235 filed Mar. 15, 2013 and titled “CRDM DESIGNS WITH SEPARATE SCRAM LATCH ENGAGEMENT AND LOCKING”. U.S. Provisional Application No. 61/792,235 filed Mar. 15, 2013 and titled “CRDM DESIGNS WITH SEPARATE SCRAM LATCH ENGAGEMENT AND LOCKING” is hereby incorporated by reference in its entirety into the specification of this application. This invention was made with Government support under Contract No. DE-NE0000583 awarded by the Department of Energy. The Government has certain rights in this invention. DeSantis et al., U.S. Pub. No. 2011/0222640 A1 published Sep. 15, 2011 and incorporated herein by reference in its entirety discloses (among other subject matter) a CRDM for a nuclear reactor employing a lead screw (sometimes referred to as a ball screw herein denoting specific lead screw embodiments employing ball nuts disposed between the screw and nut threadings) engaged by a motor to provide controlled vertical translation, in which a separate latch assembly connected with the lead screw latches to the lifting rod of a control rod (or to the lifting rod of a control rod assembly comprising plural control rods connected by a yoke or spider to the lifting rod). The latch is actively closed to connect the translating assembly comprising the lifting rod and the control rod(s) so that the translating assembly translates with the lead screw under control of the CRDM motor. Upon removal of the closing force, e.g. during a SCRAM, the latch opens to release the lifting rod and SCRAM the control rod(s), while the lead screw remains engaged with the CRDM motor and does not fall. In some illustrative embodiments, the latches are actively closed by cam bars that are lifted by a hydraulic piston, solenoid, or other lifting mechanism, where each cam bar is part of a four-bar linkage that moves the cam bar horizontally in response to the lifting in order to cam the latches shut. In DeSantis et al., U.S. Pub. No. 2011/0222640 A1, the four-bar linkage is arranged such that under gravity the four-bar linkage operates to move the cam bars outward so as to release the latch. By way of non-limiting illustrative example, FIGS. 1 and 2 correspond to drawing sheets 1 and 16, respectively, of DeSantis et al., U.S. Pub. No. 2011/0222640 A1. With reference to FIG. 1, an illustrative nuclear reactor vessel of the pressurized water reactor (PWR) type is diagrammatically depicted. An illustrated primary vessel 10 contains a reactor core 12, internal steam generator(s) 14, and internal control rods 20. The illustrative reactor vessel includes four major components, namely: 1) a lower vessel 22, 2) upper internals 24, 3) an upper vessel 26 and 4) an upper vessel head 28. A mid-flange 29 is disposed between the lower and upper vessel sections 22, 26. Other vessel configurations are also contemplated. Note that FIG. 1 is diagrammatic and does not include details such as pressure vessel penetrations for flow of secondary coolant into and out of the steam generators, electrical penetrations for electrical components, and so forth. The lower vessel 22 of the illustrative reactor vessel 10 of FIG. 1 contains the reactor core 12, which can have substantially any suitable configuration. The illustrative upper vessel 26 houses the steam generator 14 for this illustrative PWR which has an internal steam generator design (sometimes referred to as an integral PWR design). In FIG. 1, the steam generator 14 is diagrammatically shown. In a typical circulation pattern the primary coolant is heated by the reactor core 12 and rises through the central riser region 32 to exit the top of the shroud 30 whereupon the primary coolant flows back down via the downcomer region 34 and across the steam generators 14. Such primary coolant flow may be driven by natural convection, by internal or external primary coolant pumps (not illustrated), or by a combination of pump-assisted natural convection. Although an integral PWR design is illustrated, it is also contemplated for the reactor vessel to have an external steam generator (not illustrated), in which case pressure vessel penetrations allow for transfer of primary coolant to and from the external steam generator. The illustrative upper vessel head 28 is a separate component, but it is also contemplated for the vessel head to be integral with the upper vessel 26. While FIG. 1 illustrates an integral PWR, in other embodiments the PWR may not be an integral PWR, that is, in some embodiments the illustrated internal steam generators may be omitted in favor of one or more external steam generators. Still further, the illustrative PWR is an example, and in other embodiments a boiling water reactor (BWR) or other reactor design may be employed, with either internal or external steam generators. With reference to FIG. 2, a control rod system embodiment is described, e.g. suitably part of the upper internals 24 of the nuclear reactor of FIG. 1, which provides electromagnetic gray rod functionality (i.e. continuously adjustable control rod positioning) and a hydraulic latch system providing SCRAM functionality (i.e. in an emergency, the control rods can be fully inserted in order to quickly quench the nuclear reaction, an operation known in the art as a SCRAM). The control rod system of FIG. 2 allows for failsafe SCRAM of the control rod cluster without scramming the lead screw. A motor/ball nut assembly is employed, such that a lead screw 40 is permanently engaged to a ball-nut assembly 42 which provides for axial translation of the lead screw 40 by driving a motor 44. The illustrative motor 44 is mounted on a standoff 45 that positions and bottom-supports the motor 44 in the support structure of the upper internals 24; other support arrangements are contemplated. A control rod cluster (not shown) is connected to the lead screw 40 via a lifting/connecting rod or lifting/connecting rod assembly 46 and a latch assembly 48. The lead screw 40 is substantially hollow, and the lifting/connecting rod 46 fits coaxially inside the inner diameter of the lead screw 40 and is free to translate vertically within the lead screw 40. The latch assembly 48, with spring loaded latches, is attached to (i.e. mounted on) the top of the lead screw 40. When the latches of the latch assembly 48 are engaged with the lifting rod 46 they couple the lifting/connecting rod 46 to the lead screw 40 and when the latches are disengaged they release the lifting/connecting rod 46 from the lead screw 40. In the illustrated embodiment, latch engagements and disengagements are achieved by using a four-bar linkage cam system including two cam bars 50 and at least two cam bar links 52 per cam bar 50. Additional cam bar links may be added to provide further support for the cam bar. When the cam bars 50 move upward the cam bar links 52 of the four-bar linkage also cam the cam bars 50 inward so as to cause the latches of the latch assembly 48 to rotate into engagement with the lifting/connecting rod 46. In the illustrated embodiment, a hydraulic lift assembly 56 is used to raise the cam bar assemblies 50. In an alternative embodiment (not illustrated), an electric solenoid lift system is used to raise the cam bar assemblies. When a lift force is applied to the cam system, the upward and inwardly-cammed motion of the cam bars 50 rotates the latches into engagement thereby coupling the lifting/connecting rod 46 to the lead screw 40. This causes the control rod cluster to follow lead screw motion. When the lift force is removed, the cam bars 50 swing down and are cammed outward by the cam bar links 52 of the four-bar linkage allowing the latches of the latch assembly 48 to rotate out of engagement with the lifting/connecting rod 46. This de-couples the lifting/connecting rod 46 from the lead screw 40 which causes the control rod cluster to SCRAM. During a SCRAM, the lead screw 40 remains at its current hold position. After the SCRAM event, the lead screw 40 is driven to the bottom of its stroke via the electric motor 44. When the lift force is reapplied to the cam system via the hydraulic lift assembly 56, the latches of the latch assembly 48 are re-engaged and the lifting rod 46 is re-coupled to the lead screw 40, and normal operation can resume. Other latch drive modalities are contemplated, such as a pneumatic latch drive in which pneumatic pressure replaces hydraulic pressure in the illustrated lift assembly 56. In FIG. 2, the lead screw 40 is arbitrarily depicted in a partially withdrawn position for illustration purposes. The latching assembly 48 is attached to (i.e. mounted on) the top of the lead screw 40. The ball nut 42 and motor 44 are at the bottom of the control rod drive mechanism (CDRM), the latch cam bars 50 extend for the full length of mechanism stroke, and the hydraulic lift system 56 is located at the top of the mechanism. In some embodiments, the CRDM of FIG. 2 is an integral CDRM in which the entire mechanism, including the electric motor 44 and ball nut 42, and the latching assembly 48 are located within the reactor pressure vessel 10 (see FIG. 1) at full operating temperature and pressure conditions. Further illustrative embodiments of CRDM designs employing the cam bars with four-bar linkages are described in DeSantis et al., U.S. Pub. No. 2011/0222640 A1, which is incorporated herein by reference in its entirety. In some illustrative embodiments, a control rod drive mechanism (CRDM) comprises: a lead screw engaged by a CRDM motor; a lifting rod supporting at least one control rod; latches secured to the lead screw and configured to latch an upper end of the lifting rod to the lead screw; a latch engagement mechanism configured to close the latches onto the upper end of the lifting rod; and a latch holding mechanism configured to hold the latches closed; wherein the latch holding mechanism is separate from the latch engagement mechanism. In some embodiments the CRDM further comprises a four-bar linkage including cam bars, the four-bar linkage configured to drive the cam bars inward to cam the latches closed responsive to operation of the latch engagement mechanism, the latch holding mechanism configured to hold the cam bars in the inward position to keep the latches closed. In some such embodiments the four-bar linkage is configured to bias the latches closed under force of gravity. In some embodiments the latch engagement mechanism operates responsive to lowering the latches over the upper end of the lifting rod and is not effective to keep the latches closed when the latches are raised again after the latch engagement mechanism operates. In some illustrative embodiments, a control rod drive mechanism (CRDM) comprises: a lead screw engaged by a CRDM motor; a lifting rod supporting at least one control rod; latches secured to the lead screw and configured to latch an upper end of the lifting rod to the lead screw; a latch engagement mechanism configured to close the latches onto the upper end of the lifting rod; and a latch holding mechanism configured to hold the latches closed; wherein the latch engagement mechanism is not effective to keep the latches closed when the latches are supporting the weight of the lifting rod and supported at least one control rod. In some embodiments the latch holding mechanism is not effective to close the latches. In some embodiments the CRDM further comprises a four-bar linkage including cam bars, the four-bar linkage configured to drive the cam bars inward to cam the latches closed responsive to operation of the latch engagement mechanism, the latch holding mechanism configured to hold the cam bars in the inward position to keep the latches closed. In some such embodiments the four-bar linkage is configured to bias the latches closed under force of gravity. In some embodiments the latch engagement mechanism operates responsive to lowering the latches over the upper end of the lifting rod and is not effective to keep the latches closed when the latches are raised again after the latch engagement mechanism operates. In some illustrative embodiments, a control rod drive mechanism (CRDM) comprises: a lead screw engaged by a CRDM motor; a lifting rod supporting at least one control rod; latches secured to the lead screw and configured to latch an upper end of the lifting rod to the lead screw; and a four bar linkage including cam bars, the four bar linkage configured to drive the cam bars inward to cam the latches closed responsive to operation of a latch engagement mechanism; wherein the four bar linkage is configured to bias the latches closed under force of gravity. In some illustrative embodiments, a control rod drive mechanism (CRDM) includes: a CRDM motor; an element translated under control of the CRDM motor; a latch configured to latch a lifting rod supporting at least one control rod with the element translated under control of the CRDM motor; a latch engagement mechanism configured to close the latch onto the lifting rod; and a latch holding mechanism, separate from the latch engagement mechanism, configured to hold the latch in its closed position. In some illustrative embodiments, a control rod drive mechanism (CRDM) includes: a CRDM motor; an element translated under control of the CRDM motor; a latch configured to latch a lifting rod supporting at least one control rod with the element translated under control of the CRDM motor; and a four bar linkage including cam bars, the four bar linkage configured to cam the latches closed responsive to operation of a latch engagement mechanism; wherein the four bar linkage is configured to bias the latches closed under force of gravity. In some illustrative embodiments, a control rod drive mechanism (CRDM) is configured to latch onto the lifting rod of a control rod assembly and includes separate latch engagement and latch holding mechanisms. In some illustrative embodiments, a control rod drive mechanism (CRDM) is configured to latch onto the lifting rod of a control rod assembly and includes a four-bar linkage closing the latch, wherein the four-bar linkage biases the latch closed under force of gravity. Disclosed herein are improvements upon CRDM designs of DeSantis et al., U.S. Pub. No. 2011/0222640 A1 employing the cam bars with four-bar linkages. In one aspect, the CRDM is improved by separating the latch engagement and latch holding functions. This may entail increasing the number of CRDM components since a separate latch engagement mechanism and latch holding mechanism are provided. However, it is recognized herein that this increase in parts is offset by improved energy efficiency. This is because the latch engagement is a momentary event that occurs very infrequently (possibly only once per fuel cycle). In contrast, the latch holding operation is performed over the entire fuel cycle (barring any SCRAM events). By employing separate latch engagement and holding mechanisms, the latch holding mechanism is not required to perform the relatively higher-energy operation of moving the latches from the unlatched position to the latched position. Accordingly, the latch holding mechanism can be made more energy efficient. In another aspect, the latch engagement mechanism, which no longer needs to perform the latch holding function, can be substantially improved. In one embodiment (see FIGS. 3-6), the latch engagement mechanism comprises a lower camming link built into the lower portion of the CRDM, which is engaged by the latch box or housing as it is lowered toward the lifting rod (which, due to its not currently being latched, is typically located at its lowermost position corresponding to maximum insertion of the control rods into the nuclear reactor core). The lowering latch housing engages the lower camming link which is curved and mounted pivotally so that an end distal from the end cammed by the latch housing is caused to drive the cam bars inward, into the latched position. Once in the latched position, the separate latch holding mechanism is engaged, and thereafter when the latch housing is raised by the CRDM motor and lead screw the lower camming link disengages but the latch remains closed by action of the separate latch holding mechanism. In another aspect, the latch engagement mechanism is implemented as a self-engaging cam/latch system (see FIGS. 7-18). This approach is achieved by modifying the four-bar linkage such that under gravity the four-bar linkage operates to move the cam bars inward so as to engage the latch. Similar to the latch engagement of FIGS. 3-6, this latch engagement activates upon lowering the latch housing over the upper end of the lifting rod. In the self-engaging approach, the latch is normally closed due to the four-bar linkage defaulting to moving the cam bars inward under force of gravity, and the upper end of the lifting rod includes a camming surface that cams the latch open as the latch housing is lowered over the upper end of the lifting rod. Once over the camming surface of the upper end, the latch again closes under force of gravity due to the orientation of the four-bar linkage. The separate latch holding mechanism is then activated to hold the cam bars in the inward position to keep the latch closed. Surprisingly, this embodiment is capable of reliable SCRAM even though the four-bar linkage is biasing the latch closed under gravity. This is because the four-bar linkage is designed with its links at large angles and of relatively long length so that the force necessary to open the latches against the gravitational closing bias of the four-bar linkage is quite modest. (See FIGS. 7-18 and related discussion for details). Accordingly, the weight of the translating assembly (i.e. the lifting rod and the attached control rod or rods and optional spider or yoke) is sufficient to easily overcome the closing bias of the four-bar linkage. In further disclosed aspects, various embodiments of the latch holding mechanism are disclosed. See FIG. 19 and following. In the CRDM system of FIG. 2, the lift system 56 (hydraulic as shown, or alternatively an electric solenoid) supports both latch actuation and long term engagement during hold and translational operations. In the variant embodiments described in the following, features of like functionality to the CRDM of FIG. 2 (for example, the cam bars 50 and the cam bar links 52 of the four-bar linkage) are labeled with like reference numbers. With reference to FIGS. 3-6 and with contextual reference to FIG. 2, a CRDM embodiment is described in which latch activation and long term hold/translation functions are separated, resulting in reduction of operational power requirements. The CRDM comprises a mechanically actuated latching device. FIG. 3 shows an isometric view of the CRDM with the control rod (not shown) fully inserted. FIGS. 4 and 5 show isometric and side cutaway views, respectively, with the latching device disengaged. FIG. 6 shows a side cutaway view with the latch engaged. The latching mechanism utilizes the CRDM motor 44, the lead screw 40 (e.g. threadedly engaged with the CRDM motor 44 via the ball screw 42 as shown in FIG. 2) and a latch box 102 to engage the latches 104 to the top of the connecting (i.e. lifting) rod 46. Springs 106 bias the latches 104 open. The latch box 102 and spring-biased latches 104 form a latch assembly corresponding to the latch assembly 48 of FIG. 2. In FIGS. 3-6, a mounting feature 108 is shown via which the latch box 102 is mounted to the top of the lead screw 40, but the lead screw itself is omitted in FIGS. 3-6. Similarly, only the top of the lifting rod 46 is shown in FIGS. 3-6, but it is to be understood that the lifting rod 46 extends downward as shown in contextual FIG. 2.) In this operation, the control rod or rods are initially fully inserted and the upper end of the lifting rod 46 is disengaged from the latches 104. The CRDM motor 44 is then operated to cause the lead screw 40 to translate downward, thus lowering the latch box 102 toward the upper end of the lifting rod 46. The downward force supplied by the CRDM motor 44 through the ball screw 42 moves the latch box 102 into contact with a lower camming link 110 built into a lower portion 112 of the CRDM. FIGS. 4 and 5 show isometric cutaway and side cutaway views, respectively, of the state in which the latch box 102 is just beginning to contact the lower camming link 110 at a contact area 114. As seen in FIG. 6, the continued application of motor torque forces the latch box 102 downward so as to press the lower camming link 110 downward resulting in a rotary action about a pivot point 116. This rotary action lifts and translates the cam bars 50 into the engaged position so as to cam against and close the latches 104 in the latch box 102. A separate holding mechanism (not shown in FIGS. 3-6 but embodiments of which are disclosed elsewhere in this application) keeps the cam bars 50 engaged as the latch box 102 is translated back upward after the latch engagement so as to lift the lifting rod 46 and attached control rod(s) upward. (Note that the control rods are not shown in FIGS. 3-6). This approach of the embodiment of FIGS. 3-6 separates latch activation and long term hold/translation functions of the CRDM, resulting in reduction of operational power requirements. (Again, FIGS. 3-6 illustrate only the latch activation—suitable embodiments of the long term hold/translation component are described elsewhere in this application.) The separation of latch activation and long term hold/translation functions simplifies the latching assembly making it easier to manufacture and less expensive. The mechanically actuated latching device described with reference to FIGS. 3-6 is electrically operated (assuming the lead screw 40 is driven by the electric CRDM motor 44 as per FIG. 2). In combination with an electrically operated holding mechanism (again, disclosed elsewhere in this application), this constitutes an all-electric CRDM. With reference to FIGS. 7-18, a CRDM embodiment with self-engaging cam/latch system and electromagnetic holding is described. In these CRDM embodiments, the four-bar linkage is modified such that under gravity the four-bar linkage operates to move the cam bars 50 inward so as to engage the latch. These CRDM embodiments also include a holding mechanism that only holds the latch and does not perform the engagement. With reference to FIG. 7, the CRDM is shown in combination with a control rod assembly 140 connected by the lifting/connecting rod 46 via the lead (or ball) screw 40 to the CRDM which includes the motor assembly 44, a modified cam assembly 144 (with a modified four-bar linkage) and latch assembly 148. With reference to FIG. 8, an enlarged view of the CRDM of FIG. 7 is shown, including the motor 44 mounted on the standoff 45, the cam assembly 144 with modified four-bar linkage, the latch assembly 148, and an optional position sensor 149. The illustrative CRDM also includes an electromagnet holding system 150 at the top of the cam assembly 144. With reference to FIGS. 9 and 10, which show cutaway perspective view of the CRDM in SCRAM mode (fully inserted) and in normal operating mode (translating or holding the control rods), respectively, the CRDM allows for failsafe SCRAM of the control rod (or control rod cluster) 140 without the need to SCRAM the lead screw 40. The lead screw/ball nut assembly is permanently attached to the electric motor 44 (only the top of which is visible in FIG. 9) which provides for its axial translation. The control rod cluster 140 is connected to the lead screw 40 via a connecting (i.e. lifting) rod 46 and the latch assembly 148 (see FIG. 7). As seen in FIG. 9, the lead screw 40 is hollow, and the lifting rod 46 fits inside the lead screw inner diameter (ID) and is free to translate vertically within the lead screw 40. The latch assembly, with two latches 154 (although three or more latches are contemplated), is secured to the top of the lead screw 40 by a lead screw/latch assembly coupling 156 (e.g., a latch housing mounted to the upper end of the lead screw). When the latches 154 are engaged with the lifting rod 46 they couple the lifting rod 46 to the lead screw 40 (normal operation) so that the lead screw 40 and lifting rod 46 move together. When the latches 154 are disengaged they release the lifting rod 46 from the lead screw 40 (an event referred to as SCRAM). Latch engagements and disengagements are achieved by using the four-bar linkage cam system 144 with a cam bar assembly provided for each latch including a cam bar 160 and cam bar links 162. However, unlike the embodiment of FIG. 2, in the CRDM embodiments of FIGS. 7-18 the cam bar links 162 are oriented such that when gravity causes the cam bars 160 to move downward the four-bar linkage action rotates the cam bars 160 inward thereby causing the latches 154 to rotate into engagement with the lifting rod 46. Because of this self-engaging feature, there is no action required to engage the latches 154 to the lifting rod 46 (other than operating the CRDM motor 44 to lower the latch assembly 148 over the upper end of the lifting rod 46) and there are no springs for biasing the latches 154 (compare with springs 106 of the embodiment of FIGS. 3-6). Thus, force of gravity is sufficient to cause the cam bars 160 to cam the latches 154 to engage the lifting rod 46 when the lifting rod is in its lowermost position (corresponding to the control rods being fully inserted). However, force of gravity is not capable of keeping the latches 154 engaged when the CRDM of FIGS. 7-18 is operated to lift the control rod assembly 140 via the lifting rod 46. Thus, the separate holding mechanism 150 is provided, which includes electromagnets 170 and magnetic couplers 172 each connected with the upper end of a respective one of the cam bars 160. In the embodiments described herein with reference to FIGS. 7-18, the illustrative electromagnet holding system 150 is incorporated to hold the cam bars 160, and thus the latches 154, in full engagement for long term hold and translational operations. When power is removed from the electromagnets 170 (as per FIG. 9) the weight of the translating assembly 140, 46 is sufficient to rotate the latches 154 and cams bars 160 out of engagement thereby causing the CRDM to SCRAM. (The term “translating assembly” or similar phraseology refers to the combination of the lifting rod 46 and the control rod assembly 140 including a set of control rods connected with the lifting rod 46 by a yoke or spider.) While the electromagnet holding mechanism embodiment 150 is described for illustrative purposes in FIGS. 7-18, elsewhere in this application other holding mechanism embodiments are disclosed that may be substituted for the holding mechanism 150. After the SCRAM event the lead screw 40 is driven back to the bottom of its stroke via the electric CRDM motor. As the latch assembly nears the bottom of the stroke it automatically re-engages with the lifting rod 46 by cam action against the conical surface 176 of the upper end of the connecting rod 46. The same automatic re-engagement action could also be used to re-engage in the event that a control rod becomes stuck and the SCRAM does not complete. The overall CRDM assembly is shown in FIGS. 7-8. Note that the lead screw 40 may also be referred to as a “ball screw”, which is an equivalent term when the threaded engagement employs a ball nut (that is, a threaded nut/screw coupling with ball bearings disposed in the threads). The layout of the CRDM of FIGS. 7-18 is similar to illustrative CRDMs described with reference to FIG. 2. However, in the CRDM of FIGS. 7-18 the electromagnet holding system 150 at the top of the CRDM has replaced the hydraulic (or solenoid) lift assembly 56 of CRDM embodiments of FIG. 2. FIG. 9 illustrates the CRDM of FIGS. 7-18 in full SCRAM mode with the ball screw 40 and control rod assembly fully inserted. In FIG. 9 only the upper end of the lifting rod 46 (also sometimes called a connecting rod) is visible. The reversed (as compared with embodiments of FIG. 2) cam link orientation causes the four-bar linkage action under downward gravitational weight of the cam bars 160 to rotate the cam bars 160 inward into full engagement thereby causing the latches 154 to be fully engaged with (the upper end of) the lifting rod 46 of the translating assembly. This is the normal self-engaged cam bar position with no load on the latches from the translating assembly and no electromagnet holding force applied by the electromagnet holding system 150. FIG. 10 illustrates normal CRDM operation (either long term hold mode or translation of the control rod assembly under control of the CRDM motor). For this operating condition the electromagnets 170 are powered on to hold the cam bars 160, and thus the latches 154, in full engagement so that they can carry the maximum translating assembly weight force. As seen in FIG. 10, the cam bars 160 extend above the top plate of the cam housing where the magnetic couplers 172 are attached. These couplers 172, made of 410 SS magnetic material in a suitable embodiment, complete the magnetic circuit for optimum electromagnet holding force. FIG. 11 shows the CRDM of FIGS. 7-18 at the start of SCRAM. The latches 154 have been rotated out of engagement by the downward force due to the weight of the translating assembly. The latch heels, which are in contact with the cam bars 160, push the cam bars outward thereby allowing the connecting rod to SCRAM. This action is designated by the force annotation 180 in FIG. 11. FIG. 11 shows the latches 154 in the land-on-land (LOL) position just riding over the outside diameter of the upper end of the connecting rod 46. FIG. 12A illustrates the CRDM of FIGS. 7-18 with the latches 154 and cam bars 160 in the fully disengaged position. This orientation is a non-operational position that could occur if the latches 154 are “kicked” outward by the downward movement of the translating assembly during SCRAM. Although this is a non-operational position with the self-engaged cam bar design of FIGS. 7-18, it illustrates that there is ample clearance between the inside surface of the latches 154 and the connecting rod 46 for SCRAM reliability. This is shown in the inset, FIG. 12B, where the clearance dclearance is indicated. FIGS. 13A and 13B illustrate the force balance for SCRAM operation. In FIGS. 13A and 13B, the weight of the translating assembly is denoted WTA, the force pushing the cam bars outward is denoted Fpush, and the weight of the cam bars is denoted WCam Bar. In the illustrative design, the maximum force needed to push each cam bar assembly outward for SCRAM (that is, the maximum required Fpush) is only a few pounds. This lateral force component of the cam bar assembly weight WCam Bar can be minimized by increasing the orientation angle of the cam link 162, e.g. to a minimum angle of about 70° in some calculated designs. In general, making the cam link 162 longer or at a larger angle (relative to the horizontal) reduces the maximum force needed to push out the cam bars. The minimum force available to push each cam bar 160 outward is produced by latch rotation due to the downward weight force of the translating assembly. This minimum available force is based on the translating assembly weight WTA minus worst-case assumed mechanical friction drag in the control rod channel and worst-case friction at all contact surfaces. SCRAM reliability is assured since the minimum available force Fpush for SCRAM is significantly larger than the force needed for SCRAM. Advantageously, the SCRAM is totally driven by gravity with no other loads required. FIGS. 14A and 14B illustrates the force balance for normal operation. Sufficient lateral force Fhold must be applied at the heel of each latch 154 to hold the translating assembly weight WTA for various modes of operation. In the illustrative embodiment of FIGS. 7-18, this force is provided by the electromagnet holding system 150 at the top of the CRDM. Since the cam bars 160 are self-engaged, the cam bar side load reduces the needed electromagnetic force. The minimum holding force FMag needed at the holding magnet 170 to maintain latch engagement during translation of the control rod assembly is computed based on translating assembly weight WTA plus worst-case assumed mechanical friction drag in the control rod channel. In calculated designs, there is ample holding force margin for all normal operating conditions. FIGS. 15A, 15B, and 15C illustrate isometric views of the electromagnet holding system 150 at the top of the CRDM. FIG. 15A shows the fully engaged operational configuration (power to magnet 170 either on or off), FIG. 15B the SCRAM operational configuration (power to magnet 170 off) and FIG. 15C the fully disengaged operational configuration (power to magnet 170 off). In the fully engaged mode (FIG. 15A), either with or without electromagnet holding force, the magnetic couplers 172 are seated against the electromagnet housings 170. This seat provides the inward stop for the cam bars 160 and for the latches for full operational engagement. FIGS. 16A, 16B, and 16C shows plan views corresponding to the isometric views of FIGS. 15A, 15B, and 15C. It is seen from FIGS. 16A, 16B, and 16C that for all operating modes the electromagnet holding system 150 fits well within the CRDM space envelope. FIG. 17 illustrates an enlarged cutaway view of the electromagnet holding system 150 for the fully engaged condition. The electromagnets 170 are suitably hermetically sealed by welding and potted for high temperature use inside the reactor pressure vessel. Some suitable materials for the components are as follows: for the electromagnet 170, the electromagnet housing may be alloy 625 non-magnetic material, the electromagnet core may be 410 stainless steel magnetic material, and the electromagnet winding may be 24 gauge copper wire; and the magnet couplers 172 may suitably be 410 stainless steel magnetic material. Designs with these materials are estimated to provide a calculated 310 lbs of holding force. These are merely illustrative examples, and other materials and/or design-basis holding force may be employed depending upon the reactor design. FIG. 18 illustrates the latch re-engagement action. The views are labeled: (1) top left view; (2) top middle view; (3) top right view; (4) bottom left view; (5) bottom middle view; and (6) bottom right view. After a SCRAM event, when re-engagement is desired, the ball screw is driven back to the bottom by the CRDM motor. The latches 154 automatically re-engage with the lifting/connecting rod 46 as the latching assembly reaches bottom. For this purpose, a conical cam surface 176 is incorporated into the configuration of the upper end of the connecting rod 46. As the latch assembly is driven back down, the inboard surfaces of the latches 154 slide down over the top of the connecting rod 46, being cammed open by the conical cam surface 176 against the gravitational bias toward closure driven by the four-bar linkage, until the self-engaged latches 154 snap back into the normal engagement pocket. Normal operation can then resume. The same latch auto re-engagement action, as illustrated in FIG. 18, can also be used to re-engage a control rod (or bank of control rods) that becomes stuck during SCRAM. The latch assembly is driven down over the upper end of the connecting rod 46 of the stuck rod (or rod bank) until the latches 154 snap into the normal engagement pocket. If it is desired to fully insert the rods into the reactor core (as is typically the case in the event of a SCRAM), then the latching assembly is driven downward by the ball screw and motor with the latches 154 pushing downward on the stuck rod. In that scenario, the bottom surfaces of the latches 154 contact the flat portion of the engaging pocket in the connecting rod 46. As load is applied, the eccentricity of the contact surfaces causes the latches 154 to remain engaged without any additional holding system. As the motor drives the ball screw down, the latches drive the stuck rod in. With reference to FIGS. 19-22, another holding mechanism embodiment for a CRDM is described. In this regard, FIGS. 3-6 and 7-18 illustrate embodiments in which latch activation and long term hold/translation functions are separated, resulting in reduction of operational power requirements. FIGS. 3-6 illustrate an embodiment of the latch activation, while FIGS. 7-18 illustrate an embodiment of the latch activation (the self-engaging cam/latch system) in combination with an embodiment 150 of the long term hold/translation function. FIGS. 19-22 illustrate another embodiment of the long term hold/translation function, which may be used in combination with the embodiment of FIGS. 3-6 or substituted for the holding mechanism 150 of the embodiment of FIGS. 7-18. FIG. 19 shows an isometric view of the latch hold mechanism of FIGS. 19-22 operating in conjunction with the cam assembly of FIGS. 2-6, i.e. with cam bars 50. FIGS. 20 and 21 show side view and cutaway side views, respectively, of the latch hold mechanism in its disengaged position. FIG. 22 shows a side cutaway view of the latch hold mechanism in its engaged position. The holding mechanism illustrated in FIGS. 19-22 utilizes a large electromagnet 200, coupled with a magnetic hanger 202 connected with the upper ends of the cam bars 50 by pins 204, as shown in FIG. 19. The electromagnet 200 is spaced apart from the hanger 202 by support posts 206 extending from a base plate 208 secured to (or forming) the top of the cam bar assembly 144. With the CRDM engaged by an engagement mechanism (such as that described with reference to FIGS. 3-6, in illustrative FIGS. 19-22), the electromagnet 200 is activated, causing a magnetic attraction between the hanger 202 and the electromagnet 200 that holds the hanger 202 in contact with the electromagnet 200 as shown in FIG. 22 (or, in alternative embodiments, into contact with a landing surface interposed between the electromagnet and the hanger). The raised hanger bar 202 holds the cam bars 50 in their raised (i.e. engaged) position via the pins 204. When power is cut to the electromagnet 200 the attractive force between the magnet 200 and the hanger 202 is severed, causing the hanger 200 and cam bars 50 to fall to the disengaged position shown in FIGS. 20 and 21. Pin slots 210 in the hanger 202 accommodate the lateral motion of the cam bars 50 due to the four-bar linkage. The sectional views of FIGS. 21 and 22 illustrate the copper windings 212 of the electromagnet 200. By separating latch activation and long term hold/translation functions of the latch of the CRDM, it is recognized herein that the operational power requirements can be reduced, since the holding mechanism is not required to actually lift the cam bars, but merely maintains the cam bars in the lifted position after the (different) engagement mechanism operates. The separation of features simplifies the holding feature making it easier to manufacture and less expensive. With reference to FIGS. 23-32, another holding mechanism embodiment for a CRDM is described, which may be used in combination with the embodiment of FIGS. 3-6 or substituted for the holding mechanism 150 of the embodiment of FIGS. 7-18. FIGS. 23-25 show two isometric views and a plan view, respectively, of the holding mechanism in the fully engaged position. FIGS. 26-28 show two isometric views and a plan view, respectively, of the holding mechanism in the SCRAM position. FIGS. 29-31 show two isometric views and a plan view, respectively, of the holding mechanism in the fully disengaged position. The isometric view of FIGS. 23, 26, and 29 show the top region of the CRDM including the holding mechanism at a viewing angle of approximately 45°. The isometric view of FIGS. 24, 27, and 30 show the top region of the CRDM including the holding mechanism at a more oblique viewing angle than 45°. FIG. 32 illustrates a plan view of the holding mechanism with annotations of the electromagnet holding force FElect for applying a force FCam Bar sufficient to hold the cam bars 160. The holding mechanism of FIGS. 19-28 utilizes horizontal holding arms 230 that have slots 232 into which pins 234 at the tops of the cam bars 160 (e.g. cam bar pins 234) fit. When the cam bars 160 are moved to the engaged position by an engagement mechanism (e.g. such as the one described with reference to FIGS. 3-6, or the self-engaging cam/latch system of the embodiment of FIGS. 7-18), the cam bar pin 234 in each pin slot 232 pushes the holding arm 230 to rotate to a point where it is in close proximity with an electromagnet 240. The rotation is about an arm pivot point 242, and the various components of the holding mechanism are mounted on a baseplate 244 that is secured to (or forms) the top of the cam bar assembly 144. When power is applied to the electromagnets 240 they attract and hold the arms 230 which are made of magnetic material. The restrained arms, in turn, hold the cam bars 160 in the engaged position via the cam bar pins 234 in the pin slots 232 and thereby maintain latch engagement. FIGS. 23-25 shows two alternative isometric views and a top view, respectively, of the holding mechanism in this fully engaged position. With reference to FIGS. 26-28 (SCRAM mode) and FIGS. 29-31 (fully disengaged mode), when power is cut to the electromagnets 240, the attractive force between the electromagnets 240 and the arms 230 is severed, allowing the arms 230 to rotate out of engagement. The weight of the translating assembly is sufficient to disengage the latches and move the cam bars 160 away (i.e. outward) for SCRAM. During this action, the holding arms 230 freely move out of the way. With particular reference to FIG. 32, the holding mechanism of FIGS. 23-32 provides a mechanical advantage due to the configuration of the holding arms 230. This is accomplished by the relative positions of the arm pivot point 242, the cam bar contact point (i.e. the engagement between the cam bar pin 234 and the pin slot 232) and the electromagnet holding force contact point (corresponding to the location of the electromagnet 240), suitably quantified by the distance dmag between the magnet 240 and the pivot point 242 and the distance dpin between the cam bar contact point (approximately the cam bar pin 234) and the pivot point 242. Because of this mechanical advantage, the holding force FElect provided by the electromagnets 240 can be reduced to provide a given force FCam Bar for holding the cam bars 160. This facilitates the use of smaller, less complex electromagnets as the electromagnets 240, as well as lower power demands for operation. The configuration of the electromagnetic holding mechanism of FIGS. 23-32 will vary somewhat depending on the configuration of the cam bars 160 and the four bar linkage. The pin slot 232 is arranged to accommodate the horizontal cam bar travel while providing the appropriate engagement to rotate the horizontal holding arms 230. In a variant embodiment, magnets are embedded into the holding arms to provide added holding strength. In some embodiments, this added force is expected to be enough to enable the holding mechanism of FIGS. 23-32 to perform both the engagement and holding operations, and could, for example, be used in place of the hydraulic lifting assembly 56 of the embodiment of FIG. 2. By way of review, FIGS. 23-25 show the cam bars 160 and holding arms 230 in the fully engaged position, either held by the electromagnets 240 or engaged by an outside means (e.g. such as the one described with reference to FIGS. 3-6, or the self-engaging cam/latch system of the embodiment of FIGS. 7-18) prior to powering the electromagnets 240. FIGS. 26-28 show the SCRAM mode, in which the arms 230 and thus the cam bars 160 have moved sufficiently for the latches to completely release the connecting (i.e. lifting) rod and control rod assembly. FIGS. 29-31 show the fully disengaged position. Due to the 4-bar linkage action, the cam bars 160 rise and fall as they are moved laterally from engaged to disengaged positions. This action is best seen in the isometric view of FIGS. 24, 27, and 30. Since the holding arms 230 pivot about fixed support posts (the pivot arm points 242), the pin slots 232 are incorporated into the holding arms 230 to accommodate the rise and fall of the cam bars 160. These slots 232 should be sized and positioned to accommodate both the rise and fall of the cam bars 160 and the lateral motion of the cam bars 160 due the four-bar linkage action responding to the rise/fall of the cam bars 160. When used in conjunction with the self-engaging cam/latch system described herein with reference to FIGS. 7-18, the direct mechanical advantage for the illustrated locations of the holding arm pivot points 242 has been estimated to be approximately 4.5:1 (corresponding to the ratio dmag/dpin in FIG. 32). However, there is not a direct relationship between this mechanical advantage and the holding force needed since the holding arms 230 do not pull in line with the plane of collapse of the cam bars 160. A force correction is needed that is proportional to the cosine of the holding arm angle. The net effect for the configuration shown herein is an effective mechanical advantage of 2.4:1. This force balance, along with the effective mechanical advantage, is diagrammatically illustrated in FIG. 32. The holding mechanism of FIGS. 23-32 has the benefit of a mechanical advantage provided by the configuration of the holding arms. With reference to FIGS. 33-38, another holding mechanism embodiment for a CRDM is described, which may be used in combination with the embodiment of FIGS. 3-6 or substituted for the holding mechanism 150 of the embodiment of FIGS. 7-18. FIGS. 33-35 show two isometric views at different viewing angles and a top view, respectively, of the top of the CRDM (and more particularly the top of the cam assembly and the holding mechanism) with the cam system in the unlatched position. FIGS. 36-38 show two isometric views at different viewing angles and a top view, respectively, of the top of the CRDM including the holding mechanism with the cam system in the latched position. Illustrative FIGS. 33-38 show the holding mechanism in combination with the embodiment of FIGS. 3-6, and hence the cam bars are labeled cam bars 50 in FIGS. 33-38. Once the cam system is in the engaged (i.e. “latched”) position the holding mechanism of FIGS. 33-38 holds the cam bars 50 such that they engage the latches and maintain latching of the connecting (i.e. lifting) rod. The holding mechanism of FIGS. 33-38 includes two high temperature magnets 260 and magnetic links 262 attached to the upper end of each of the two cam bars 50 at the top end of the CRDM. The two canned high temperature electromagnets are suitably wired in a parallel fashion. When the cam system transitions from the unlatched position (FIGS. 33-35) to the engaged (latched) position (FIGS. 36-38), the upper ends of the cam bars 50 engaging the magnetic links 262 rotate the magnetic links 262 about pivots 264 so that the ends 270 of the magnetic links 262 distal from the cam bar/magnetic link joint 272 are moved by the inward movement of the cam bars 50 to be in close proximity to the electromagnets 260. When the electromagnets 260 are energized these distal ends 270 of the magnetic links 262 are held against the magnets 270, and the cam bar 50 at the opposite end of the link 262 is prevented from moving. This holds the latch in the latched position. The holding power of the electromagnets 260 is adequate to hold the weight of the cam bars 50 as well as the force exerted on the cam bars 50 by the latches. The latched state is shown in alternative isometric views (FIGS. 36 and 37) and a plan view (FIG. 38). Slots 276 in a base plate 278 secured to (or forming) the top of the cam bar assembly and supporting the hold mechanism components accommodate the lateral motion of the cam bars 50 during unlatched/latched transitions. When used in conjunction with the embodiment of FIGS. 3-6 (as illustrated in FIGS. 33-38), operation is as follows. When the electromagnets 260 are de-energized the magnetic links 262 are decoupled from the electromagnets 260 and the cam bars 50 are free to fall under their own weight and swing into the unlatched position. In the unlatched position the cam bars 50 are disengaged from the latches and the latches can then rotate out of engagement with the connecting rod. When the cam bars 50 are disengaged from the latches, the latches can be rotated out of engagement with the connecting rod by the latch springs 106 (for the embodiment of FIGS. 3-6). Therefore, in the unlatched position the cam bars 50 are not engaged with the latches, the latches are not engaged with the lifting rod and the translating assembly (including the lifting rod and the attached control rod or rods) can then fall under its own weight (SCRAM). The holding mechanism of FIGS. 33-38 is fail-safe in the sense that if power is lost to the electromagnets 260 the connecting rod will SCRAM due to gravity. Operation of the holding mechanism of FIGS. 33-38 in conjunction with the cam arrangement of FIGS. 7-18 (self-latching) is similar, except that when the electromagnets 260 are de-energized the cam bars 160 do not open under gravity, but rather are cammed open by the cam surface at the upper end of the lifting rod 46 of the falling translating assembly. (See description of FIGS. 7-18 for details). Again, the de-energizing of the electromagnets 260 allows the magnetic links 262, and hence the cam bars 160, to freely move to perform the SCRAM. With reference to FIGS. 39-48, another holding mechanism embodiment for a CRDM is described, which may be used in combination with the embodiment of FIGS. 3-6 or substituted for the holding mechanism 150 of the embodiment of FIGS. 7-18. The embodiment of FIGS. 39-48 is illustrated in conjunction with a four-bar linkage with cam bars and cam bar links oriented as in the embodiments of FIGS. 2-6; accordingly, in FIGS. 39-48 the cam bars and cam bar links are labeled as cam bars 50 and cam bar links 52, respectively. The embodiment of FIGS. 39-48 illustrates a variant latching mechanism located beneath the cam assembly, in which a hydraulic cylinder 300 (or, alternatively, an electric solenoid) raises a lift plunger or piston 302 upward to engage cam bar lift rollers 304 at the bottom ends of the cam bars 50 so as to raise the cam bars 50—by action of the four-bar linkage provided by cam bar links 52 this raising of the cam bars 50 simultaneously moves the cam bars 50 inward to engage the latch. (By comparison, in the embodiment described with reference to FIG. 2, the hydraulic lift assembly 56 located above the cam assembly lifts the upper ends of the cam bars 50 to engage the latches). The embodiment of FIGS. 39-48 also illustrates a holding mechanism located above the cam assembly, where a base plate 308 secured to (or forming) the top of the cam bar assembly supports the hold mechanism components. FIG. 39 shows a diagrammatic side view of the cam assembly, in which the lift system (comprising electric solenoid or hydraulic cylinder 300 and piston 302 in conjunction with cam bar lift rollers 304) is deactivated and the hold mechanism (diagrammatically shown in a tilted view) is also deactivated. FIG. 40 shows a top view of the deactivated hold mechanism corresponding to FIG. 39. FIG. 41 shows a diagrammatic side view of the cam assembly in which the lift system is activated and the hold mechanism is still deactivated. FIG. 42 shows a top view of the deactivated hold mechanism corresponding to FIG. 41. FIG. 43 shows a diagrammatic side view of the cam assembly in which both the lift system and the hold mechanism are activated, and FIG. 44 shows a corresponding top view of the activated hold mechanism. FIG. 45 shows a diagrammatic side view of the cam assembly in which the lift system is deactivated and the hold mechanism is still activated, and FIG. 46 shows a corresponding top view of the activated hold mechanism. FIGS. 47 and 48 illustrate geometric aspects of the hold mechanism. The hold mechanism of the embodiment of FIGS. 39-48 keeps the four-bar linkage cam system 50, 52 in the engaged position during rod translation and hold functions, and provides the SCRAM functionality when subsequently deactivated. It also structurally internalizes the majority of the cam bar retention force required to hold the latches in the engaged position, and utilizes mechanical advantage to minimize the remaining hold force, resulting in a structurally efficient unit. FIGS. 39 and 40 illustrate the holding mechanism (and associated lift system in FIG. 39) both in the deactivated state. The holding mechanism including a rotary hold bar 310, a hold-solenoid 312 (where the housing of the solenoid 312 is visible), a hold-solenoid plunger 314, and hold-bar rollers 316, is located at the top or base plate 308 of the cam bar assembly. FIGS. 39 and 40 illustrate the hold mechanism deactivated at startup. Prior to startup, the lift system (electric solenoid or hydraulic), which includes the electric solenoid or hydraulic cylinder 300 and the lift plunger or piston 302, is also deactivated. Therefore, the latches are not engaged by the four-bar cam system 50, 52, rendering the connecting rod and attached control rods in the fully inserted position. As best seen in the top view of FIG. 40, in the unlatched state of the four-bar linkage 50, 52 the cam bars 50 are in their outboard positions (i.e., moved outward and away from the latches). Also note that the base plate 308 includes slots to accommodate movement of the upper ends of the cam bars 50 between their inboard (i.e. moved in) and outboard (i.e. moved out) horizontal positions. With reference to FIGS. 41 and 42, upon activation of the lift system (shown in FIG. 41), the lift plunger or piston 302 raises the cam bars 50 into the latch engagement position by contact with the cam bar lift rollers 304. At initial engagement of the lift mechanism, the hold mechanism is still deactivated as depicted in FIGS. 41 and 42. Because of activation of the lift system, the latches are now engaged with the connecting rod which is resting with the attached control rods at the fully inserted position. As best seen in FIG. 42, the lifting of the cam bars 50 also moves the cam bars 50 into their inboard positions by action of the four-bar linkage, and this inward movement is what engages the latches, as described in more detail with reference to the embodiments of FIGS. 2-6. With reference to FIGS. 43 and 44, subsequently following activation of the lift system, the hold solenoid 312 of the hold mechanism is activated, resulting in extension of the solenoid plunger 314, which rotates the hold bar 310 about a pivoting engagement 318 of the hold bar 310 with the base plate 308. At full extension of the solenoid plunger 314, the hold-bar rollers 316 are rotated into position behind the upper extremity (i.e. upper ends) of the cam bars 50 (note again that the upper ends of the cam bars 50 protrude through the slots in the base plate 308), so as to function in the hold capacity. It is noted that the hold solenoid 312 is free to pivot about a post mount 320 that secures the solenoid 312 on the base plate 308. It is also noted that the solenoid plunger 314 is pin-connected to the hold bar 310, which provides rotational freedom for operation. The relative orientations of all the pertinent components at this phase of operation are illustrated in FIGS. 43 and 44. With reference to FIGS. 45 and 46, with the hold mechanism activated the lift system can be deactivated, with the hold system thereafter keeping the latches engaged. Upon deactivation of the lift system, the lift plunger or piston 302 is released, and therefore, no longer (bottom) supports the cam bars 50. At this point, the cam bars 50 are retained in the engaged position solely by the hold mechanism. The four-bar cam system 50, 52 is now being retained for long-term retention of the connecting rod by the hold mechanism. With reference to FIG. 47, there exists an eccentricity Econtact between the center of rotation of the hold bar 310 and the line of action of the contact force between (the upper end of) the cam bar 50 and the hold-bar roller 316. This eccentricity Econtact results in a force-moment imbalance on the hold bar 310 when the force applied by the hold solenoid 312 is removed. This moment imbalance at power loss to the hold solenoid 312 is the driving mechanism for rapidly rotating the hold bar 310 and the attached rollers 316 out of contact with the cam bars 50—resulting in SCRAM (rapid release of connecting rod). In order to create a smooth rolling action of the hold-bar rollers 316 on the contact surface of the cam bars 50, the contact surface is contoured to the arc of the rolling-contact point. With continuing reference to FIG. 47 and with further reference to FIG. 48, the desired lower power consumption of the hold mechanism is a product of the significant mechanical advantage of the unit. The moment arm Eplunger of the hold solenoid plunger 314, relative to the pivot center of the hold bar 310, is significantly larger than the moment arm of the contact force of the cam bar 50 at the hold-bar roller 316, as illustrated in FIGS. 47 and 48. Therefore, the force required by the hold solenoid 312 is significantly less than the latch-to-cam bar contact force required to support the connecting rod load. Of further advantage, internalization of the majority of the cam bar retention forces as equal and opposite loads reacted through the hold bar 310 eliminates force reaction through the remainder of the hold mechanism, resulting in a structurally efficient unit. As previously stated, the hold mechanism described with reference to FIGS. 39-48 separates latch activation and long term hold/translation functions, resulting in reduction of operational power requirements. The hold mechanism keeps the four-bar linkage cam system in the engaged position during rod translation and hold functions, and provides the SCRAM functionality when subsequently deactivated. It also structurally internalizes the majority of the cam bar retention force required to hold the latches in the engaged position, and utilizes mechanical advantage to minimize the remaining hold force, resulting in a structurally efficient unit. With reference to FIGS. 49-52, another holding mechanism embodiment for a CRDM is described, which may be used in combination with the embodiment of FIGS. 3-6 or substituted for the holding mechanism 150 of the embodiment of FIGS. 7-18. FIG. 49 shows an isometric view of the top region of the CRDM including the holding mechanism with the vertical linkage engaged to raise the cam bars. FIG. 50 shows a corresponding isometric view with the vertical linkage disengaged to allow the cam bars to fall. FIG. 51 corresponds to the engaged view of FIG. 49 but includes a partial cutaway, and similarly FIG. 52 corresponds to the disengaged view of FIG. 50 but includes the partial cutaway. The latch holding mechanism of FIGS. 49-52 utilizes a vertical linkage system including vertical links 340 connected to a hanger 342 disposed between (the upper ends of) the cam bars 160 of FIGS. 7-18 (as shown; or, alternatively, the cam bars 50 of FIGS. 2-6) and (in the engaged position shown in FIGS. 49 and 51) held in the engaged position by electromagnets 344. When the cam bars 160 are moved to the engaged position by the separate latch engagement mechanism (e.g. as in the embodiment of FIGS. 3-6, or the embodiment of FIGS. 7-18), it causes the hanger 342 to move up which, in turn, raises the vertical links 340 to a position where horizontal drive members 348 are in close proximity with the electromagnets 344. When power is applied to the electromagnets 344 they attract and hold magnets that are embedded into the horizontal drive members 348. (Alternatively, the horizontal members 348 may be made of steel or another ferromagnetic material but not include magnets). The restrained vertical links 340, in turn, hold the hanger 342, and thus the cam bars 160, in the engaged position and thereby maintain latch engagement. When power is cut to the electromagnets 344, the attractive force between the electromagnets 344 and the horizontal drive members 348 is severed, allowing the vertical links 340 to drop out of engagement, as seen in FIGS. 50 and 52. The weight of the translating assembly is sufficient to disengage the latches and move the cam bars 160 away for SCRAM. During this action, the linkage system freely moves downward out of the way. To recap, FIGS. 49 and 50 show isometric views of the top region of the CRDM at a viewing angle of approximately 45° for the engaged and disengaged states, respectively. FIG. 49 shows the vertical linkage system in the fully engaged (full up) position, either held by the electromagnets 344 or engaged by an outside means prior to powering the electromagnets. For the SCRAM mode, shown in FIG. 50, the linkage system has moved full down for the latches to completely release the connecting rod and control rod assembly. FIGS. 51 and 52 show isometric cutaway views of the top region of the CRDM for the engaged and disengaged states, respectively. FIG. 51 shows the vertical linkage system in the fully engaged (full up) position, either held by the electromagnets 344 or engaged by an outside means prior to powering the electromagnets 344. FIG. 52 shows the linkage system in the full down (SCRAM) position. In the illustrative embodiment, the minimum angle of the vertical links 340, in the fully engaged position (FIGS. 49 and 51), is set to about 10° which is expected to assure an adequate SCRAM reliability margin. In the disengaged position (FIGS. 50 and 52) the vertical links 340 collapse to a maximum angle of about 40° in the illustrative embodiment. The latch holding mechanism described with reference to FIGS. 49-52 provides a mechanical advantage due to the configuration of the linkage system. This is due to the relative positions and size of the vertical link 340 lengths compared to the horizontal drive member 348. In addition, the permanent magnet that is embedded in the horizontal arm 348 provides added holding force. The true mechanical advantage for this disclosed vertical linkage system is calculated to be 2.9:1 at the minimum link angle. However, the effective mechanical advantage is higher, estimated to be closer to 4.0:1, when an assumed permanent magnet force per link assembly is added. Because of this mechanical advantage, the required holding force needed by the electromagnets is reduced. This results in smaller, less complex electromagnets, as well as lower power demands for operation. Illustrative embodiments including the preferred embodiments have been described. While specific embodiments have been shown and described in detail to illustrate the application and principles of the invention and methods, it will be understood that it is not intended that the present invention be limited thereto and that the invention may be embodied otherwise without departing from such principles. In some embodiments of the invention, certain features of the invention may sometimes be used to advantage without a corresponding use of the other features. Accordingly, all such changes and embodiments properly fall within the scope of the following claims. Obviously, modifications and alterations will occur to others upon reading and understanding the preceding detailed description. It is intended that the present disclosure be construed as including all such modifications and alterations insofar as they come within the scope of the appended claims or the equivalents thereof.
	abstract	An X-ray exposure apparatus operating in a vacuum or a reduced-pressure environment includes a mask, arranged in the vacuum or in the reduced-pressure environment, for holding a reflection X-ray mask having a mask pattern thereon, an X-ray illuminating system arranged so as to illuminate the reflection X-ray mask having the mask pattern thereon, relatively scanningly with X-rays, wherein the mask pattern being illuminated is transferred onto an object, and a cooling structure for cooling the X-ray mask held by the mask chuck.
	description	This invention generally relates to electron probe micro-analysis (EPMA). More particularly, this invention relates to analyzing the properties, such as thickness and composition, of a sample, such as layers used to form integrated circuits on a semiconducting substrate. Integrated circuit fabrication is typically accomplished by forming many different layers on a substrate. As used herein, the phrase integrated circuit refers to circuits such as those formed on monolithic substrates of a semiconducting material, such as group IV materials like silicon and germanium, and group III-V compounds such as gallium arsenide. Because the design tolerances of an integrated circuit are so strict, it is desirable to monitor the properties, such as thickness and elemental composition, of the various layers as they are formed. One way to measure the properties of film layers is to use electron microprobe x-ray spectrometry. Electron microprobe x-ray spectrometry uses an electron beam source to excite a sample. X-rays having wavelengths that are characteristic of the elements of the sample are emitted from the sample over a continuous range of takeoff angles, defined as the angle between the x-ray and the sample surface. An x-ray detector assembly is positioned to detect a fraction of the x-rays that are emitted from the sample. The detector assembly can capture x-rays emitted over a finite range of takeoff angles. The detector assembly includes both a spectrometer and an x-ray detector. The spectrometer selects x-rays within a narrow range of wavelengths and directs only those x-rays to the x-ray detector. This is typically accomplished by rotating a diffractor through a range of angles, where at each angular position of the diffractor, the diffractor deflects x-rays with a given wavelength range towards the detector. The rate of impingement of the x-rays within subsets of the desired range of wavelengths is sequentially detected and measured. From this information, properties such as the elemental composition and thickness of the sample can eventually be determined. There are several methods of detecting the x-rays with different energies. A first method uses a set of curved crystal or multilayer reflectors attached to a rotatable turret. As used herein, a “turret” refers to a carousel-like holder of one or more objects, such as reflectors, disposed in a circumferential relationship to each other with respect to some axis of rotation and positions the objects at a particular location by rotation of the holder about the axis. One reflector is positioned so that it reflects and focuses x-rays through an aperture into a gas proportional counter. The gas proportional counter converts the x-ray into an electrical pulse that is detected by an electronic detection system. Each of the reflectors on the turret can reflect and focus x-rays over a fixed Bragg energy range. At a particular position and orientation, a first reflector can reflect and focus x-rays over a first narrow energy range. This first narrow energy range is of the right size to capture the characteristic x-rays emitted by a first element. The first reflector can be rotated and repositioned to reflect x-rays emitted from a second element contained within a second narrow range within the first reflector's Bragg range. In order to capture x-rays from a third element not contained with the first reflector's Bragg range, it is necessary to rotate the turret to bring a second reflector into position. This reflector can reflect and focus the x-rays from a third element contained within a third narrow energy range and within the second reflector's Bragg range. The union of all the reflectors' Bragg energy ranges determines how many elements can be detected. If enough detectors are included, and if they have overlapping Bragg energy ranges, it is possible to detect almost all elements in the periodic table. In addition, it is often desirable to measure the x-rays with energies on either side of the narrow range emitted from the element being measured. This can be accomplished by periodically rotating and repositioning the reflector by small amounts, and recording the number of x-rays at each position. In other words, the detector is scanned in energy. In this way the x-ray spectrum can be measured in the energy region around and including the element's narrow range. This is useful for determining the background x-ray intensity that is later subtracted from the elemental narrow range measurement to get the true x-ray intensity emitted from the element. There are several drawbacks to the above method. For example, in order to capture the x-rays with high efficiency it is desirable for the reflector to intercept the x-rays with as large a solid angle as possible. A large solid angle for collecting x-rays allows measurements to be made quickly. This requires either positioning the reflector close to the x-ray source, or using a very large area reflector. Unfortunately, the turret design places geometrical restrictions on the number of large solid angle reflectors that can be accommodated by the turret. Commercially available systems typically accommodate from two to a maximum of six reflectors. It is desirable to have a larger number of reflectors, both to be able to cover a large number of elements and to optimize the efficiency for each element. Prior art x-ray detection systems suffer from other drawbacks. For example, in order to have a reflector with a moderately large Bragg energy range, or to be able to scan it to measure a spectrum, it is necessary for it to be curved in a circular cylindrical shape. An inside surface of the cylindrical reflector has multilayer dielectric coating having a d-spacing that is constant over a surface of the dielectric. The reflector, source (e.g., sample surface) and detector are located at points on a circle known as the Rowland circle such that the detector stays at a linear focus of the reflector. The Rowland circle has a diameter that is half the diameter of the cylindrical reflector. Unfortunately, a circular cylindrical reflector is not an optimum shape to achieve the best efficiency for a particular element. In addition, the solid angle for collecting x-rays is limited to about 0.03 steradians. Attempts have been made to improve the efficiency of cylindrical reflectors. One technique uses a multilayer d-spacing having a gradient across the reflector to keep the efficiency high. Unfortunately, the optic cannot be scanned with much efficiency and the solid angle for collection of x-rays is limited to about 0.10 steradians. Another technique uses a three-dimensional ellipsoidal multilayer optic that can focus x-rays from a point source to a small spot image. Unfortunately, the ellipsoidal shape is difficult and expensive to manufacture. In addition. the ellipsoidal multilayer optic also uses a multilayer dielectric with a graded d-spacing. Consequently, the optic cannot be scanned. What is needed, therefore, are large collection angle x-ray monochromators that can overcome some of the problems described above. Embodiments of the present invention are related to turret-less x-ray monochromators and electron probe micro-analysis (EPMA) systems using such monochromators. A first embodiment of the invention relates to a turretless x-ray monochromator. The monochromator may use a cassette of reflectors instead of a turret. The cassette stores a plurality of reflectors that can be selectively inserted in situ into a conventional Rowland circle monochromator geometry. A transfer mechanism selectively moves reflectors from the cassette to a positioning mechanism. The use of the cassette allows each reflector to be placed closer to a source of x-rays, thereby allowing a larger solid angle for x-ray collection. A second embodiment of the invention uses a non-focusing reflector that can be fixed, scanned axially or scanned radially to provide large solid angle detection of x-rays at various energies with a single reflector. Although the following detailed description contains many specific details for the purposes of illustration, anyone of ordinary skill in the art will appreciate that many variations and alterations to the following details are within the scope of the invention. Accordingly, the exemplary embodiments of the invention described below are set forth without any loss of generality to, and without imposing limitations upon, the claimed invention. As illustrated, e.g., in FIG. 1 and FIG. 2, certain embodiments of the present invention use a cassette of reflectors instead of a turret. FIG. 1 depicts an x-ray monochromator 100 according to a first embodiment of the present invention. The monochromator includes a cassette 102 adapted to retain multiple x-ray reflectors 104 and an x-ray detector 105. A positioning mechanism 106 selects a reflector 104 from the cassette 102 and rotates it into position for reflection of x-rays from a known location 108 toward the detector 105. With the cassette 102, one or more x-ray reflectors may be selectively positioned in situ in close proximity to the known location 108. As used herein, the term “selectively positioned in situ” refers to placing an object into position in a closed chamber without having to open the chamber. It will be seen from the following discussion that the use of the cassette allows the reflectors to be selectively positioned in situ in close proximity to the known location 108 without the use of a turret. In the example depicted in FIG. 1, the positioning mechanism, positions the x-ray reflectors such that have a reflecting surface that intersects a Rowland circle 107. The detector 105 and known location may also lie on the Rowland circle 107. The cassette 102 could be placed at a large enough distance from known location 108 that it would not interfere with placing the reflector 104 close enough to the known location to collect a large solid angle (e.g., greater than about 0.010 steradians. The set of reflectors 104 could include more than 5 reflectors, and is not limited by the geometrical constraints of the x-ray source region. By way of example, and without limitation, the reflectors 104 may include one or more scannable reflectors, e.g., having a circular cylindrical symmetry about an axis perpendicular to the plane of the drawing in FIG. 1. Alternatively, the reflectors 104 may include one or more elliptical cylindrical reflectors or one or more ellipsoidal reflectors. The reflectors 104 may have a multilayer coating characterized by an interlayer spacing d (referred to herein as the d-spacing). The multilayer coating may be made of alternating layers of metal and carbon or metal and silicon or different metals, e.g., chromium and scandium. Each layer may be about 10 angstroms thick. The interlayer spacing d of the coating determines the Bragg angle θB with respect to a plane of the coating for constructive interference of the x-rays. The condition for constructive interference is 2d sin θB=mλ, where, m is an integer and λ is the x-ray wavelength. The x-ray wavelength may be related to the x-ray energy E by E=hc/λ, where h is Planck's constant and c is the speed of light. The coating may have a d-spacing with a gradient such that the d-spacing varies across the surface of the reflector 104. The d-spacing may vary along a circumference of the reflector 104, in a direction perpendicular to the axis of curvature. The d-spacing of the multilayer coating may be selected for different energy wavelength ranges. Different materials for the multilayer coating may be used along with different d-spacings, in order to optimize performance for different wavelength ranges. For example, a tungsten-silicon multilayer dielectric may have a 2d-spacing of about 60 angstroms. A chrome-scandium multilayer dielectric may be used to produce a d-spacing of about 80 angstroms. The set of reflectors 104 may also include non-scannable reflectors highly optimized for individual elements. These non-scannable detectors could have one or more of the following properties: multilayer coatings that are not efficient at more than one narrow energy range, multilayer coatings that have a graded interlayer spacing d for increased solid angle collection, elliptical cylindrical shapes for improved focusing and background rejection, and doubly curved ellipsoidal shapes for very large solid angle collection and improved background rejection. By way of example, the reflectors 104 may include between 7 and 20 different reflectors with a different reflector for every K, L and M x-ray line between about 0.15 keV and about 2.0 keV. Such a set of reflectors would allow detection of K-lines for boron, carbon, nitrogen, oxygen, sodium, and fluorine, L-lines for cobalt, copper, gallium, arsenic, nickel and zinc, among others. The reflectors 104 are placed into position by the reflector positioning mechanism 106. This would also scan the reflector during a spectrum measurement. The mask containing the aperture and the detector are mounted on a base plate 109. This base plate is positioned and scanned by the positioning mechanism 106. The cassette 102 is moved vertically by the cassette positioner 114. This allows the reflector 104 to always be picked up or replaced at the same location by a transfer mechanism 116. By way of example, the transfer mechanism 116 may be in the form of a reflector pusher used to push the reflector 104 out of the cassette 102 into the reflector positioning mechanism 106 with a simple linear motion. A detector positioner 118 may be used to position the detector 105, e.g., to keep the detector on the Rowland circle 107. The x-ray detection system 100 may include a vacuum chamber 110 that houses the x-ray detector 105. The x-ray detector 105 may have an entrance aperture 112 that restricts the entry of x-rays into the detector 105. The cassette 102, reflectors 104 and positioning mechanism 106 may be situated within the vacuum chamber 110. In some circumstances, it may be desirable to have more than one mask aperture 112 available for use with the x-ray monochromator 100. For example circular cylindrical reflectors focus to a wide linear spot, elliptical cylindrical reflectors focus to a narrower linear spot, and ellipsoidal reflectors focus to a small circular spot. In order to get the optimum background rejection for each type of reflector, it is best to use an aperture only slightly larger than the focal spot and of the same shape. More than one aperture could be mounted on an aperture selector mechanism 118, which in turn is mounted on the detector base plate 109. The aperture selection mechanism may selectively place different apertures in front of the detector 105. The vacuum chamber 110 may be attached to the source of x-rays, which may be an EPMA system 200 as depicted in FIG. 2. The EPMA system 200 includes an electron beam column 202 that directs a beam of electrons 211 onto a surface of a sample 204. An interaction between the sample and the electrons 211 generates x-rays 205. The electron beam column 202 and sample 204 may be housed within a vacuum chamber 210. The vacuum chamber 110 containing the x-ray monochromator 100 is attached to the EPMA system 200 so that an entry aperture 207 is aligned as an x-ray detector port for the EPMA system 200. The vacuum chambers 110, 210 may be provided with sealing flanges to ensure a good vacuum seal. In this way, the invention would replace one of the standard detectors on an EPMA system. An example of a suitable EPMA system is the Metrix 100 available from KLA-Tencor Corporation of San Jose, Calif. Any suitable type of x-ray detector may be used as the x-ray detector 105 in x-ray monochromator of FIG. 1 and the EPMA system of FIG. 2. By way of example, the detector 105 may be a dual chamber gas proportional counter 300 as illustrated in FIG. 3. The counter 300 includes a first chamber 302 containing a first gas 304 and a first anode 306. X-rays 301 enter the first chamber 302 through a first window 308 and initiate a discharge to the first anode 306 though interactions with the first gas 304. Some of the x-rays entering the first chamber 302 may subsequently enter a second chamber 312 having a second gas 314 and second anode 316. A second window 318 allows x-rays to enter the second chamber 312 from the first chamber 302. The use of two different gases 304, 314 and windows 308, 318 made of different materials allows the detector 300 to detect x-rays in different energy ranges. By way of example, the first gas 304 may be neon gas, the second gas 314 may be xenon gas, the first window 308 may be made of a polymer and the second window 318 may be made of beryllium. Alternatively, a combination of xenon and carbon dioxide gases may be used. In certain embodiments, the detector 105 may be a position sensitive detector having an array of two or more detector elements 115 as indicated in FIG. 2. Each element, e.g., a charge-coupled-device, semiconductor detector, etc., may be located at a different location. X-rays emerging at different angles from the known location 108 on the sample 204 strike the detector 105 at different positions. Thus, the position of detection can provide angular information about the x-rays. The angular information, in turn, can provide information about the sample, 204. For example, the sample may include a copper layer overlying tantalum nitride layer disposed on a silicon substrate. In this example, there are three variables of interest, which are the thickness of the tantalum nitride layer, the nitrogen concentration of the tantalum nitride layer, and the thickness of the copper layer. Without the angle resolved detection the EPMA system 200 would generate only three data values, which are the copper x-ray counts, the tantalum x-ray counts, and the nitrogen x-ray counts. With such limited data, it is generally quite difficult to distinguish variations in the nitrogen concentration of the tantalum nitride layer from thickness variations in the copper layer, because the copper layer tends to absorb the nitrogen x-rays as they are emitted from the tantalum nitride layer. Thus, the data tends to be confounded between at least two possible causes of variation. However, nitrogen x-ray absorption produces an angle dependent intensity. Nitrogen x-rays with higher takeoff angles, or in other words with takeoff angles that are closer to perpendicular to the surface of sample 204, will tend to be more intense because the nitrogen x-rays have traveled through less of the copper layer thickness as they escaped the sample. This angular variation in x-ray intensity provides additional information in regard to both the thickness of the copper layer and the concentration of the nitrogen in the tantalum nitride layer. The reflector 104, e.g., a parabolic collimator is preferably adapted to work best for low energy x-rays like nitrogen, and a 12 degree range of collected takeoff angles is enough to detect a significant variation in x-ray intensity with takeoff angle using a position sensitive detector as the detector 105. The above embodiments of the present invention overcome a tradeoff between efficiency and flexibility. The above embodiments enable detection of a wide range of x-ray energies and a large number of elements, while maintaining optimum detection efficiency for each. It also does all this while taking up the space of a single x-ray detector on an EPMA system, leaving room for multiple standard detectors on the other ports. In the above embodiments the reflector selection mechanism is removed from the x-ray source region in order to maintain a large solid angle of x-ray collection. The cassette-based design also facilitates combining both scannable and non-scannable reflectors in a single detector system. Alternative embodiments of the present invention use a reflector with a shape that does not focus the x-rays. Dropping the requirement of focusing allows much greater freedom in choosing the shape and it can be optimized for various combinations of solid angle collection, ease of manufacture, and scanning. The non-focusing optic requires a larger detector than a focusing detector, due to its diffuse output. This detector can be either a gas-filled proportional counter (e.g., as shown in FIG. 3) or a solid state detector. In either case, the detector can be in the form of an array and can determine the position of the x-ray on the detector in order to determine the angle of the incoming x-ray as described above. By way of example and without limitation, FIG. 4 depicts an x-ray monochromator 400 according to an alternative embodiment of the present invention. In this example, the monochromator 400 includes an x-ray optic 402 in the shape of a cylinder. X-rays 401 emerge from a source located on a central axis z of the cylinder. An inside surface 404 of the optic 402 has a multilayer coating 406. A d-spacing d of this coating may be graded to maintain the correct Bragg angle along the length of the optic 402. A central region of the cylindrical optic 402 is blocked by a disc 408 of x-ray absorbing material to prevent unreflected x-rays and stray electrons from entering a detector 410. The optic 402 may be in a fixed positional relationship with respect to the source of x-rays 401. By way of example, in an EPMA, the x-rays 401 may be produced at a surface of a sample 412 as a result of interaction with a beam of electrons 411 from an electron column 414. The detector 410 may be an array of detectors, as described above, to allow for position sensitive angle-dependent detection of x-rays of different take-off angles from the source, as described above. A maximum length L of the optic without double reflection is 2w, where w is a working distance between a front end of the optic and the known location of the source of x-rays (e.g., the sample surface in the case of an EPMA system). Double reflection of the x-rays by the optic 402 is not desirable since the second reflection typically has nearly zero efficiency due to the graded multilayer coating 406. In practice the optic 402 may be manufactured in two or more sections, split along the axis, in order to allow the multilayer coating to be deposited or otherwise formed on the inside of the cylinder. It is possible to construct a radial scanning version of the monochromator 400 if the cylindrical optic 402 is fabricated such that it has a variable radius R. If the radius R is increased, the optic 402 can reflect x-rays emerging at angles with respect to the surface normal. The larger the angle, the smaller the energy of the x-rays. Thus, for a larger radius R, x-rays of a smaller energy will be reflected to the detector. If the original angle of incidence is not too large, x-rays of approximately the same energy will be reflected at all positions of the cylindrical optic 402. In other words, the reflectivity response versus energy will shift, but not broaden significantly. This is just what is desired for a scanning monochromator. The blocking disc 408 preferably moves forward as the cylinder radius R increases in order to prevent unreflected x-rays from passing through to the detector. FIGS. 5A-5B depict an example of a radially scanning x-ray optic 500 that may be used as the optic 402 in the monochromator 400. In this example the optic 500 is made in multiple cylindrical segments 502 that can move radially with respect to a common central axis. Each segment may include on an interior surface thereof a multilayer dielectric coating, which may have a d-spacing that varies along the axis. As a practical matter, each segment 502 may be made of a substantially rigid material that does not change its curvature as they move radially. As a result, when the array expands from an initial radius R0 to an expanded radius R, the reflecting surface of the expanded array of cylindrical segments 502 would not be circular. The non-circularity could introduce additional broadening, but increasing the number of segments limits this effect. FIG. 6 depicts a plot of broadening versus energy shift for different numbers of segments. A dashed line 602 indicates broadening where a cylindrical optic of fixed radius is scanned axially with respect to the source of x-rays. A dotted line 604 indicates broadening where a four-segment cylindrical optic is scanned radially. A solid line 606 indicates broadening for a six-segment cylindrical optic that is scanned radially. From these graphs it can be seen that it is desirable to have at least six (6) segments to keep the broadening to an acceptable level. Other non-focusing shapes are possible for the optic 402. For example, FIG. 7A depicts a conical non-focusing optic 700 having a reflector surface 702 in the shape of a truncated cone and an x-ray blocking central disc 704. The optic 700 can collect even larger solid angles than a cylindrical design. Although such a design cannot be practically scanned, it is useful for angle sensitive detection, as described above. A more complicated horn-like optic 710 is shown in FIG. 7B. The optic 710 includes a reflector 712 having a horn-like shape and a central stop 714. An advantage of such a design is that the reflector can be scanned in the axial direction (i.e., by moving the optic 712 along a symmetry axis z of the reflector 712). Such a configuration may be more practical than a radial scan of a cylinder. The horn-like shaped reflector 712 has the shape of a surface of revolution based on the curve ƒ(z) as shown in FIG. 8. An entrance aperture 716 of the optic 712 is located at the plane z=0, and the reflective surface is located at the radius r=ƒ(z). A point source of x-rays is located on the axis of symmetry at z=−w0 so that the nominal working distance is w0. In practice, the reflector 712 may be scanned by translating it along the axis of symmetry (the z-axis), thereby changing the working distance. For the purpose of analysis, it is easier to think of keeping the reflector fixed and translating the source so that the working distance is a variable equal to a nominal value w0 plus a small variable scanning distance a.w=w0+a The reflector will usually be scanned over a small range, 2a0, centered on the nominal working distance w0. In the discussion and equations that follow the quantity a0 is a fixed parameter for the total scanning range of the variable working distance was follows:w0−a0≦w≦w0+a0 The entrance aperture of the reflector subtends a half angle θ0, as shown in the diagram. In this case an x-ray from the source reflects from the surface at the point (z, ƒ(z)) and makes an angle, θz with respect to the z-axis. θ z = tan - 1 ⁡ [ f ⁡ ( z ) z + ( w 0 + a ) ] The angle of the surface, θs, is equal toθs=tan−1 ƒ″(z) The x-ray makes an angle of incidence, θinc, with the surface of θz−θs. θ inc = θ z - θ s = tan - 1 ⁡ [ f ⁡ ( z ) z + ( w 0 + a ) ] - tan - 1 ⁢ f ′ ⁡ ( z ) In order to be reflected efficiently, the angle of incidence θinc must equal the Bragg angle for the multilayer film stack at that point. Since the angle of incidence varies along the surface, the multilayer film must also change accordingly. This is accomplished by varying the spacing of the film layers along the surface so that the Bragg angle matches the angle of incidence at each point of the surface for the desired x-ray energy. The correct layer spacing, d, can be calculated from the Bragg equation, where λ is the desired x-ray wavelengthλ=2d sin θinc This can be written in the more useful form d ⁡ ( z ) = hc E 0 ⁢ sin ⁢ ⁢ θ inc ⁡ ( z ) = hc E 0 ⁢ sin ⁡ [ tan - 1 ⁡ [ f ⁡ ( z ) z + w 0 ] - tan - 1 ⁢ f ′ ⁡ ( z ) ] where h is Planck's constant, c is the speed of light, E0 is the energy of the desired x-ray and ƒ′(z) is the derivative of ƒ(z) with respect to z. If the reflector is manufactured so that the film layer spacing varies along the surface according to this equation, x-rays with energy E0 will be reflected efficiently from each part of the surface. This will be true for any shape of surface, i.e., for any function ƒ(z), as long as the reflector is at the nominal position (w=w0). It is noted that the above equation may be used to determine the gradient of the d-spacing for the simple case of a circularly cylindrical non-focusing reflector for which ƒ(z)=R, where R is a constant radius of a cylinder. However, the efficiency of reflectivity for such a cylinder would not remain constant as the cylinder scans axially in the z direction. The problem becomes to choose a surface shape so that this property of efficient reflectivity remains at least approximately true when we scan the reflector. If the reflector moves by a distance a, the new working distance is w=w0+a. The angles of incidence are now different than they were at the nominal position, and the Bragg condition will be satisfied by a different x-ray energy. Consider two different points along the surface at z1 and z2. With the reflector at the nominal position they have two angles of incidence θinc1 and θinc2. The layer spacing d is designed so that they both reflect a single energy E0. After moving the reflector a distance a, the two points will have two new angles of incidence θinc1′ and θinc2′. There will be two new x-ray energies that will be efficiently reflected, E1′ and E2′. These x-ray energies can be calculated by the equations E 1 ′ = E 0 ⁢ sin ⁢ ⁢ θ inc ⁢ ⁢ 1 sin ⁢ ⁢ θ inc ⁢ ⁢ 1 ⁢ ′ E 2 ′ = E 0 ⁢ sin ⁢ ⁢ θ inc ⁢ ⁢ 2 sin ⁢ ⁢ θ inc ⁢ ⁢ 2 ′ It is desirable to have these two energies equal to each other, for any two points on the surface z1 and z2. We can see that if we make the ratio of the sines of the angles independent of z, this goal will be reached. For an arbitrary shape, the ratio will depend on both a and z, but we want to find a shape to minimize or eliminate the variation with z. The ratio determines the ratio of the new reflected energy to the nominal desired energy. We call this the scanning function, h(a,z). h ⁡ ( a , z ) = E ′ ⁡ ( a , z ) E 0 = sin ⁢ ⁢ θ inc ⁢ ⁡ ( z ) sin ⁢ ⁢ θ inc ′ ⁡ ( a , z ) = sin ⁡ [ tan - 1 ⁡ [ f ⁡ ( z ) z + w 0 ] - tan - 1 ⁢ f ′ ⁡ ( z ) ] sin ⁡ [ tan - 1 ⁡ [ f ⁡ ( z ) z + w 0 + a ] - tan - 1 ⁢ f ′ ⁡ ( z ) ] One method of finding a good surface shape is to start with a trial shape and vary it to minimize the variation of h (a, z) with z. For instance, a possible trial function ƒ1(z) is a 3rd order polynomial.ƒ1(z)=b0+b1z+b2z2+b3z3 It is possible to construct a function which measures the variation with z by evaluating the difference of h(a, z) from h(a, 0) at N points (z1, z2, z3, . . . , zN) along the surface and summing the squares of the errors. The worst variation will probably occur for a=a0, so it may be best to construct the function for that case. The resulting variation, v0, can be calculated for any trial function. v 0 = ∑ n = 1 n = N ⁢ ( h ⁡ ( a 0 , z n ) - h ⁡ ( a 0 , 0 ) ) 2 It may be advantageous to construct a set of such variations, each evaluated at a different value of a (a0, a1, a2, a3, etc.) and then sum the squares of these vn for an improved measure of the variation, v. In either case, it only remains to vary the parameters of the trial function (b0, b1, b2, b3 in this case) until the variation is minimized. Minimization problems of this type are well known in the art, and various algorithms exist to solve them. One commonly used algorithm is known as Levenburg-Marquardt. Another method for finding a shape is to require that the scanning function be strictly independent of z for a particular value of a, say a0. We can do this by differentiating h with respect to z and setting the derivative equal to 0. ⅆ h ⁡ ( a 0 , z ) ⅆ z = 0 This produces an implicit second order differential equation for ƒ(z). If a definite value of a is inserted, this equation can be solved using any of a number of well-known numerical methods to produce a shape ƒ(z) which has zero variation of reflected energy along the surface for the particular value of a used in the solution, as well as for a=0. The variation in energy for other values of a will not be zero. The equation may be solved several times, each with a different value of a, and the various shapes obtained may be averaged to yield a shape in which the variation is not too large for all values of a to be included in the scan. One well-known method of numerically solving differential equations is known as Runge-Kutta. Still another method of finding a shape makes use of the fact that the scanning function, h(a, z), depends linearly on a for values of a that are much smaller than the nominal working distance (a<<w0). We can expand the expression for h in a power series in a and keep only the constant and linear term. h ⁡ ( a , z ) = 1 - af ⁡ ( z ) ⁢ ( w 0 + z + f ⁡ ( z ) ⁢ f ′ ⁡ ( z ) ) ( ( w 0 + z ) 2 + f ⁡ ( z ) 2 ) ⁢ ( ( w 0 + z ) ⁢ f ′ ⁡ ( z ) - f ⁡ ( z ) ) + O ⁡ [ a 2 ] = 1 + ga + O ⁡ [ a 2 ] We want to find a shape to make h independent of z. We can do that by forcing g to be a constant independent of z, which generates a first order differential equation for ƒ(z). This equation can be solved by the same techniques mention earlier for the second order equation. f ′ ⁡ ( z ) = f ⁡ ( z ) ⁢ ( w 0 + z ) ⁡ [ gf ⁡ ( z ) 2 + g ⁡ ( w 0 + z ) - 1 ] g ⁡ ( w 0 + z ) 3 + f ⁡ ( z ) 2 ⁡ [ g ⁡ ( w 0 + z ) + 1 ] The solution of this equation produces a family of solutions parameterized by the constant g. These solutions allow efficient reflection over the entire reflector as long as the scan distance, a, is much smaller than the nominal working distance, w0. A non-focusing multilayer optic of the types described herein may be used as monochromators in an x-ray detection sub-assembly of an EPMA. Radially or axially scannable non-focusing x-ray monochromators as described herein may obviate the need for multiple reflectors thereby allowing for a turret-less monochromator while retaining the energy scanning capabilities of a multiple system. The freedom of shape possible with non-focusing optics allows for one or more of larger collected solid angle, scanability, and angular detection. Use of non-focusing x-ray optics in EPMA systems would enable higher precision and faster throughput of EPMA applications to thin film measurement in semiconductor fabrication. Such embodiments of the present invention would enable the background level around an emission line to be measured and therefore increase the accuracy and matching of EPMA applications. Such embodiments of the present invention would also aid in the measurement of concentration depth profiles by the use of emission angle information. Non-focusing x-ray monochromators of the type described herein may be used in conjunction with a cassette configuration as described with respect to FIG. 1 and FIG. 2. In addition, x-ray monochromators that use non-focusing reflectors as described herein may be used in conjunction with a more conventional turret design. Turret-less x-ray monochromators and EPMA systems according to embodiments of the present invention provide numerous advantages over conventional turret-based designs. Cassette-based embodiments of the present invention allow for greater solid angle collection than in turret-based designs by overcoming geometric constraints placed on the working distance in a turret-based design. Non-focusing embodiments of the present invention enable the collection of much larger solid angles of x-rays than previous cylindrical designs. In a fixed configuration, non-focusing monochromators may be of lower cost and easier to manufacture than an ellipsoidal design. Scannable versions of non-focusing monochromators enable scanning where the ellipsoidal design does not. Any of the embodiments can be used with a position sensitive detector to obtain information about the emission angles of the x-rays. The position sensitive detector may be a semiconductor array, charge-coupled-device (CCD), multiwire proportional counter or other position sensitive x-ray detector. While the above is a complete description of the preferred embodiment of the present invention, it is possible to use various alternatives, modifications and equivalents. Therefore, the scope of the present invention should be determined not with reference to the above description but should, instead, be determined with reference to the appended claims, along with their full scope of equivalents. In the claims that follow, the indefinite article “A”, or “An” refers to a quantity of one or more of the item following the article, except where expressly stated otherwise. The appended claims are not to be interpreted as including means-plus-function limitations, unless such a limitation is explicitly recited in a given claim using the phrase “means for.”
060289066	abstract	The present invention relates to a control rod installed in a boiling water reactor (BWR), and particularly to a control rod using metal. An object of the present invention is to provide a manufacturing method of a control rod for a boiling water reactor, having excellent corrosion resistance in high temperature water and excellent wear resistance at the time of fabrication, and in which the influence of the manufacturing process is slight. In order to achieve the above described object, the present invention provides a manufacturing method of a control rod for a boiling water reactor constructed with sheaths having a U-shaped cross section attached to each end of a tie rod having a cruciform cross section, and rod, plate or oval tube cross section metal hafnium type neutron absorber material contained inside the sheaths, in which an anodized film is provided on a surface of the neutron absorber material as a pre-process of assembly of the neutron absorber material in the structure of the control rod.
	description	This application claims the benefit of U. S. Provisional Application Nos. 62/394,311 filed Sep. 14, 2016 and 62/394,631 filed Sep. 14, 2016 both of which are herein incorporated by reference. Not applicable. Not applicable. For remote outposts and isolated and small communities, securing a fossil fuels supply could be costly and challenging. Small, portable nuclear reactor power systems are a practical and superior alternative for providing reliable electrical power supply and/or process heat to these communities and disaster relief efforts for extended periods without refueling. The development of small reactors for terrestrial power has been explored, more than 60 years ago, with current interest in many countries. Developing portable compact reactor power systems has also been of military interest. As shown in Table I, various military units have differing power requirements. TABLE 1Electrical Power Requirements of US military forward operating basesElectrical Power Requirements (kWe)BattalionBrigadePerCompany (600(3500person(150 people)people)people)CENTCOM Sand 0.7105 4202450Book“Base in a Box” concept1.827010806300Air Force Expedition 1.36750 (5504500 Airfieldpeople)(3300people) The U.S. Army's Nuclear Power Program has produced the truck-portable ML-1 design, tested in the Gas-Cooled Reactor Experiment, and the modular PM-1, PM-2, and PM-3 pressurized power reactor (PWR) designs, deployed to Greenland, Ak., and Antarctica. The development of mobile and compact reactor concepts takes into account several important requirements. To enable mobility and rapid transport and deployment, the reactor and the power system need to have a long-life and be lightweight and compatible with a variety of common modes of transportation. In addition to using a qualified transportation cask with protective barriers against impact and radiological release, in the unlikely event of a transportation accident, the reactor design should provide for maximum nuclear safeguards. For enhanced performance and reliability of the reactor and the power system may consider passive means for cooling, such as natural convection without pumps, low operating pressure, static or a hybrid static and dynamic energy conversion options with ambient air cooling. Other considerations include avoidance of single point failures in reactor design and power system integration, load-following and maximum utilization of the reactor thermal power. For example, rejecting residual heat to ambient air enhances mobility and allows the power system operation and deployment in arid regions. This heat could be used for space heating in cold regions and water desalination in coastal regions. The LPORTS could be installed at a selected site below ground and mounted on seismic insolation bearings to avoid impact by projectiles and vulnerability to Earthquakes. In one embodiment, the present invention provides power systems that may be factory manufactured, assembled and sealed, for maximum nuclear safeguards and highest quality control, and shipped to site by rail, on large truck or on a barge, for maximum mobility. In other embodiments, the present invention provides a Long-life, Portable and compact nuclear Reactor for efficient Terrestrial power Systems (LPORTS) capable of generating 100 kWe-1.0 MWe, for 10-20 full power year (FPY), or even longer, without refueling. The reactor design takes advantage of the good thermal properties and fission gas retentions, and compatibility of UN fuel with conventional CrFe steel cladding and core structure. The load-following, factory fabricated and sealed reactor provides for maximum nuclear safeguards and reliable passive operation. Natural circulation of in-vessel liquid sodium, with the aid of a tall in-vessel chimney and a Na/Na heat exchanger (HEX), cools the reactor core during nominal operation and after shutdown. In other embodiments, the LPORTS of the present invention may use liquid metal heat pipes conductively coupled to the reactor vessel wall, or an in-vessel helically coiled tubes Na/Na HEX, near the top of the down-comer. In other embodiments, the LPORTS of the present invention may use an ex-vessel sliding reflector or rotating drums for a redundant reactor control. The near atmospheric operation reduces the wall thickness and both the volume and mass of the reactor vessel. Also, the in-vessel sodium provides effective energy storage of decay heat after reactor shutdown, maintaining a large safety margin from boiling temperature. Furthermore, the sodium, in a frozen state, provides structural support of the core and structure components during transportation. In other embodiments, the LPORTS of the present invention, for energy conversion, may use a thermally regenerative electrochemical device for the direct conversion of heat to electrical energy. In other embodiments, the LPORTS of the present invention use as the electrochemical device a static Alkali Metal Thermal-to-Electric Conversion with a thermoelectric bottom cycle (AMTEC-TE). The system may be cooled by natural circulation of ambient air to be used for space heating, and use combined dynamic cycle to generate electricity, for a total system thermal efficiency ˜60% and total utilization of the reactor thermal power>80% In other embodiments, a plurality of modular AMTEC-TE units may be connected thermal-hydraulically and electrically in parallel to provide for maximum redundancy and the avoidance of single point failures. In other embodiments, the LPORTS of the present invention may be configured to use process heat for the co-generation of alternative fuels, and low-grade heat for space heating and seawater desalination. In other embodiments, the LPORTS of the present invention may use biological shielding materials and configurations, and options of integrating the reactor into a reliable, low maintenance power system for rapid removal and deployment and transport, while adequately shielded in a qualified impact resistant container or cask. Detailed embodiments of the present invention are disclosed herein; however, it is to be understood that the disclosed embodiments are merely exemplary of the invention, which may be embodied in various forms. Therefore, specific structural and functional details disclosed herein are not to be interpreted as limiting, but merely as a representative basis for teaching one skilled in the art to variously employ the present invention in virtually any appropriately detailed method, structure or system. Further, the terms and phrases used herein are not intended to be limiting, but rather to provide an understandable description of the invention. In one embodiment, the LPORTS of the present invention may use one or more thermally regenerative electrochemical devices for the direct conversion of heat to electrical energy. A preferred device is a static Alkali Metal Thermal-to-Electric Conversion with a thermoelectric bottom cycle (AMTEC-TE), cooled by natural circulation of ambient air, and combined dynamic cycle for a system total thermal efficiency ˜60%. As shown in FIG. 1A, LPORTS 100 includes reactor 102, an Alkali Metal Thermal-to-Electric Conversion (AMTEC) 110 that provides modular unit designs and a wide range of electric powers from a kWe to a MWe. The AMTEC thermal efficiency is the highest fraction of Carnot of any dynamic and other static conversion technologies known today. Potassium AMTEC units, designed and analyzed for generating 6.0 kWe to 1.0 MWe, typically operate at a hot side temperature of 950-1050 K and a condenser temperature of 500-600 K. At these temperatures, residual heat is directed to bottom TE elements 120, which generated attritional electricity, reject waste heat to ambient air using water heat pipes fins 125. The primary working fluid exits HEX 128 in the AMTEC units, connected thermal-hydraulically in parallel, at 900-1000 K. This enables the co-production of alternative fuels and/or co-generation of high voltage AC electricity using combined dynamic energy conversion of a top closed Brayton cycle 130 and a bottom Rankine steam cycle 140, with air cooling as shown in FIG. 1. In an alternate embodiment, as shown in FIG. 1B, AMTEC 110 may be coupled by HEX 160 to Rankine steam cycle 140, with air cooling. In another alternate embodiment, as shown in FIG. 1C, AMTEC 110 may be coupled by HEX 162 to closed Brayton cycle 130. In yet another alternate embodiment, as shown in FIG. 1D, reactor 170 may be coupled to Rankine steam cycle 172. The (AMTEC-TE) units may also be configured to provide electrical power at 200-400 VDC, or lower, at a thermal efficiency of 25-30%. The combined cycle, also cooled by ambient air, will provide high voltage AC electricity at a thermal efficiency of 45-50%, for an overall power system efficiency more than 60%. The parallel-connected AMTEC-TE units provide for maximum redundancy and allow replacing malfunctioning units without shutting down the reactor. As shown in FIG. 2, in another embodiment the present invention provides a Safe InterModal Portable Long-life Energy reactor power system (SIMPLE). The SIMPLE reactor power system 200 combines a high temperature sodium cooled reactor 210, cooled by the natural circulation of the in-vessel sodium, with a plurality of thermally regenerative electrochemical devices 220-222 for the direct conversion of heat to electrical energy which may be AMTEC conversion units with integrated TE bottoming cycles thermally connected to secondary liquid-sodium loops 230-232 with static electromagnetic pumps (EMP) 240-242 and heat rejection by the natural circulation of ambient air. This embodiment provides a robust plant configuration that is entirely passive and inherently load following, with no moving mechanical parts and redundant reactor control mechanisms. Reliability may be enhanced by multiple secondary loops 240-242 which allow the system to continue operation in the event of a failure in one of the EMPs or AMTEC units. Furthermore, the reactor core is contained within a sealed reactor vessel, with no duct penetrations, utilizing external reactivity control and heat removal through the vessel wall by conductively coupled heat exchangers to the secondary sodium loops. Increased energy utilization is made possible through the intermediate heat exchanger (IHX) 280 connected to the secondary sodium loops 230-232. Reactor's 210 thermal power would be increased with the additional thermal energy generation transferred through the IHX to process heat applications, such as the high-temperature production of liquid transportation fuels. Alternatively, IHX 280 may connect to bottoming cycle module, such as a superheated steam cycle plant, for additional electrical power generation. This would allow the versatile SIMPLE system to adapt to a variety of mission electricity and thermal energy demands using a common modular architecture. Further development and analyses of the SIMPLE concept are needed to develop a detailed point design which meets the safety and performance requirements within the size and mass limitations. As shown in FIG. 3, reactor power system 300 may be emplaced below grade to protect the reactor from external attack and to utilize the surrounding earth 310 as a supplement to radiation shielding 312. The SIMPLE reactor concept provides protection against the release of radioactive material in the event of a transportation accident using a strong reactor vessel and by allowing the primary sodium coolant to freeze before transport, encasing the core within a solid block. The self-contained SIMPLE power system module may be configured to be capable of supplying ˜100-300 kWe for ten years, supporting a company scale field installation. The self-contained SIMPLE reactor concept may be designed to fit within the weight limit (<30 MT) and dimensions of a high cube intermodal shipping container. This enables the design to be transported and handled using existing infrastructure and equipment, allowing for its rapid transport and delivery by standard semi-truck, rail car, ship, or military cargo aircraft. As shown in FIG. 4, in one preferred embodiment of the present invention, transportation module 400 may be used to transport reactor 405. Module 400 may include housing 410, cask 412, radiation shielding 414 and a plurality of spaced apart impact absorbing sections 425A-425C. The other components of the system 430 may also be included in the module. In a preferred embodiment, reactor 405 is located in cask 412 which is surrounded by radiation shielding 414. To reduce and/or prevent damage during transport, these components may be isolated from the housing by impact absorbing sections 425B and 425C. Impact absorbing sections 425B and 425C may partially surround the components by locating the components in-between the sections or panels of the absorber. In other embodiments, absorber surrounds the components. Housing 410 may also include a section to house the other components of the system such as other top and bottom cycle components. To reduce and/or prevent damage during transport, these components may be isolated from the housing by impact absorbing sections 425A and 425B. Impact absorbing sections 425A and 425B may partially surround the components by locating the components in-between the sections or panels of the absorber. In other embodiments, absorber surrounds the components. In yet another embodiment, the present invention is configured so that the decay heat after reactor shutdown is removed safely using passive means of natural circulation of in-vessel liquid sodium aided by liquid metal heat pipes along the wall of the reactor primary vessel as well as by natural circulation of ambient air at the outer surface of the reactor guard vessel. The large inventory of the in-vessel liquid sodium also provides excellent thermal energy storage. In addition to the redundant reactor control and emergency shutdown, the large negative temperature reactivity of the reactor core and the in-vessel liquid sodium could shutdown the reactor with modest increase in temperature, while maintaining a large temperature safety margin from the boiling temperature of sodium. In short, core meltdown is eliminated in the embodiments of the present invention. While the foregoing written description enables one of ordinary skill to make and use what is considered presently to be the best mode thereof, those of ordinary skill will understand and appreciate the existence of variations, combinations, and equivalents of the specific embodiment, method, and examples herein. The disclosure should therefore not be limited by the above-described embodiments, methods, and examples, but by all embodiments and methods within the scope and spirit of the disclosure.
059828390	abstract	Automated inspection assemblies for scanning pipes of a nuclear reactor are described. In one embodiment, the assembly includes a mounting subassembly and a scanning subassembly. The mounting subassembly includes a clamp configured to be mounted to selected pipes in a nuclear reactor pressure vessel, and the scanning subassembly is movably coupled to the mounting subassembly. The scanning assembly includes a scanning head configured to scan at least a portion of the circumference of the pipe to be inspected, and the scanning head includes a substantially "U" shaped transducer support assembly sized to receive the pipe. Transducer elements are coupled to the transducer support assembly legs, and the transducer support assembly is configured to rotate about the pipe to inspect the pipe.
	summary
	abstract	The present invention provides an X-ray examination apparatus, comprising
041535065	description	DESCRIPTION OF THE PREFERRED EMBODIMENTS Referring to FIG. 1, there is illustrated one example of a fuel power region in which an operation of a nuclear reactor is preferably started without causing pellet-cladding interactions in the fuel rods. For the type of reactor under discussion, the pellets are for example about 12.4 mm in diameter in a 7 .times. 7 array of spaced fuel rods, and 10.6 mm in a 8 .times. 8 array of spaced fuel rods. The initial diametral gap is about 0.3 mm for the 7 .times. 7 array, and 0.23 mm for the 8 .times. 8 array. The solid line shows an experimental result of an upper limit of the fuel power level, above which the periphery of the pellet expands to contact the inner surface of the cladding by the expansion of the pellets. The upper limit of the fuel power level P.sub.1 can be expressed approximately by the following equation (1). EQU P.sub.1 = 366 .times. (G/D) - 1.38 (1) in which G is an initial circumferential gap (diametral gap) between the pellet and cladding, D is the diameter of the pellet, and the dimension of of the fuel power P is Kw/ft. When the fuel power reaches a power level determined by the equation (1), the pellets deform to contact the inner surface of the claddings and is capable of causing cracks in the corrosion resistance coating on the inner surface of the cladding. According to one embodiment of the present invention, the operation of the reactor is started within a fuel power level P determined from the following inequality (2), that is, EQU P < 366 .times. (G/D) - 1.38 (2) the fuel power is kept at this power level P for a certain period of time which is long enough to dehydrate the pellet to remove the moisture contained in the fuel rods. Further, it is necessary to dehydrate the moisture as fast as possible so that the reactor can be operated at a desired maximum power level within a short period. As the dehydration of the pellets is promoted at a temperature of the surface of the pellet more than about 400.degree. C., it is necessary to increase the fuel power to heat the pellet above 400.degree. C. The single dot-dash line and the double dot-dash dotted line in FIG. 1 show the lower fuel power limits in which the surface of the pellets is heated more than 400.degree. C., respectively at temperatures of the coolant water of 240.degree. C. and 288.degree. C. Both the single and double dot-dash lines can be approximately expressed by the following equation (3), that is, EQU P.sub.2 = 9.54 - 123 .times. (G/D) - 0.016 T (3) in which T is the temperature of the coolant water surrounding the fuel rods, and P.sub.2 is a lower power limit in which the temperature of the surface of the pellets becomes greater than 400.degree. C. Therefore, it is preferable to increase the fuel power P to satisfy the following inequality; EQU P < 9.54 - 1.23 .times. (G/D) - 0.016 T (4) referring to FIG. 2, there is illustrated an experimental result of H.sub.2 O adsorbing efficiency of a getter interposed in a fuel rod, and also illustrated an experimental result without a getter. In these experiments, 60 mg of H.sub.2 O was adopted in a fuel rod and the fuel rod was circumferentially cooled at a temperature of 288.degree. C. As apparent from the solid line in FIG. 2, the residual H.sub.2 O in the fuel rod decreases to 10 mg after 15 hours heat treatment. It is known that 2 mg of H.sub.2 O per 1 mm.sup.3 of open space in the fuel rod is sufficient to cause hydride damage to the claddings. This residual amount of 10 mg H.sub.2 O is one twelfth (1/12) of the amount sufficient to cause hydride localization in the fuel rods, and therefore about 15 to 20 hours are enough to make the residual H.sub.2 O absorbed into the getter interposed in the fuel rod. Accordingly, in the starting operation of the present invention, 15 to 20 hours are sufficient for the holding time, in which the fuel power is kept between the power levels determined by the equations (1) and (3). It is preferable to make the holding time as short as possible to operate the reactor efficiently, so that, it is also preferable to operate the reactor at the highest power level within that region to release the residual H.sub.2 O from the pellets. Referring to FIG. 3, there is illustrated one embodiment of the starting operation of the present invention. In this embodiment, the diameter of the pellets is 10.6 mm, the diametral gap between the pellets and claddings is 0.23 mm and the temperature of the coolant is 288.degree. C. Therefore, the upper limit of the fuel power P.sub.1 is determined from the equation (1); P.sub.1 = 6.56 Kw/ft, and the lower limit of the fuel power P.sub.2 is also determined from the equation (3); P.sub.2 = 2.26 Kw/ft. In this embodiment, the fuel power is increased at a relatively rapid rate until about 3 Kw/ft and which power level is kept constant for about 20 hours. After 20 hours, the power level is abruptly increased to about 6 Kw/ft which amount is sufficient not to cause the interactions and is thereafter increased to a maximum desired power level at a rate of 0.06 Kw/ft/hour. It is not always necessary to maintain the starting power level P constant for the holding time, that is, the starting power level can be varied for the predetermined holding time within the power levels determined from the equations (1) and (3). A typical power plant of Boiling Water Reactor, in which the present invention can be applied, is schematically illustrated in FIG. 4. A pressure vessel 100 contains a nuclear fuel core 101 and a steam separating and drying apparatus 102. A plurality of control rods 103 may be reciprocated by drive devices 104 into and out of the core 101 to control the reactivity of the reactor. A rod control system 105 controls the operation of the control rod drive device 104. The vessel 100 is filled with a coolant to a level somewhat above the core 101. The coolant is circulated through the core 101 by a circulation pump 106 through a pipe 109 which receive the coolant from a fuel core shroud 108 and forces it into a plenum 107 from which the coolant flows upward through the fuel assemblies of the reactor core. The heat produced by the fuel elements is thereby transferred to the coolant and a head of steam is produced in the upper portion of the vessel 100. The steam is applied to the turbine 110 to drive an electric generator 113. The turbine exhausts to a condenser 111 and the resulting condensate is returned as a feed water to the vessel 100 by a feed water pump 112. A neutron detector 114 is provided in the fuel core 101 to detect the power produced by the fuel elements. The detected signal produced in the neutron detector is transferred through a circuit 115 to a calculating system 116. The calculating system 116 and the rod control system 105 are connected by a circuit 117. The calculating system 116 begins to operate according to a signal indicating the use of fresh fuel rods through a circuit 118. When a signal indicating an installation of fresh fuel rods in the reactor is applied to the calculating system 116 through the circuit 118, the calculating system 116 begins to operate in accordance with the flow chart illustrated in FIG. 5. The calculating processes in the calculating system 116 will be explained in detail with reference to FIG. 5. At a starting operation of the reactor in which fresh fuel rods are installed in the fuel core 101, a signal indicating an existence of the fresh fuel rods is manually applied through the circuit 118 to the calculating system 116. This signal input step needs to be completed at least before the starting operation of the reactor. In the calculating circuit, a judgment of the existence of new fuel rods is made, that is, when the new fuel rods are installed in the reactor, subsequent steps in the direction of "YES" are followed. On the other hand, when no new fuel rods are installed in the reactor, a conventional starting operation is employed. When the new fuel rods are installed in the reactor, the fuel power P.sub.2 is evaluated in accordance with the equation (3). The values of the inside diameter of the fuel pellet D, and the initial circumferential gap G between the pellet and the cladding are applied beforehand to the calculating system 116. The coolant temperature is detected by a temperature detector for example thermocouples (not shown) arranged in the fuel core 101 and the value of the coolant temperature is also applied as an input to the calculating system 116. After completing the evaluation of the fuel power P.sub.2, the fuel power P.sub.1 is evaluated in accordance with the equation (1). The fuel power P, which is in proportion to the output power of the reactor detected by the neutron detector 114 is raised within a power range determined by the detected values of the fuel power P.sub.1 and P.sub.2 to satisfy the following inequality; EQU P.sub.1 > P .gtoreq. P.sub.2 when the fuel power is raised up to reach P, the fuel power is held at a constant value P for a certain holding time. During the holding time, moisture contained in the fresh fuel rods is absorbed in the getter installed within the fuel rods. After completion of the absorption, the fuel power P is then raised in accordance with the conventional operating method. As is readily apparent to those skilled in the art, the type of the rod control system 105 and the calculating system 116 are chosen from structures available in the art of the nuclear power control, for example, of the type disclosed in U.S. Pat. No. 3,565,760 which describes in more detail the calculating system 116 of the present invention. As is apparent from this disclosure, the region of the fuel power level in which the starting operation is preferred is not limited to the FIG. 1 illustration or the equations (1) and (3), because the upper and the lower power limits differ in accordance with the material used as the fuel pellets or the claddings. It is to be noted that, the feature of the present invention is that, the fuel power level at the starting period of the reactor having fresh fuel rods is maintained below a power level in which no interaction appears in the fresh fuel rods. This invention can be applied even to a starting operation of a nuclear reactor in which fresh fuel rods are installed without the corrosion resistance coating on the inner surface of the claddings. In the conventional method of starting operation, the residual moisture or hydrogen included in the fuel rods would be captured in a narrow space surrounded by the contact portions of the pellets and claddings within the gap, which may cause the localized hydriding of the fuel rod. However, according to the present method, the residual moisture or hydrogen in the fuel rods is removed before the appearance of the interaction between the pellets and claddings. In the above description, a Boiling Water Reactor is employed for explanation purposes, but this invention can be applied to any other type of reactors which employ nuclear fuel pellets covered with fuel rod claddings. According to the features of the present invention, fuel rod failures can be eliminated by removing the residual moisture before the appearance of the interaction. As a result of the elimination of the fuel rod failures, a safe and efficient starting operation of the nuclear reactor can be obtained. It is understood that the present invention is not limited to the specific features described herein but also contemplates numerous changes and modifications as would be known to those skilled in the art given the present disclosure of the invention, and we therefore do not wish to be limited to the details shown and described herein only schematically, but intend to cover all such changes and modifcations.
	summary
	description	This application claims the benefit of U.S. Provisional Application No. 61/625,764 filed Apr. 18, 2012 and titled “UPPER INTERNALS”. U.S. Provisional Application No. 61/625,764 filed Apr. 18, 2012 titled “UPPER INTERNALS” is hereby incorporated by reference in its entirety into the specification of this application. This application claims the benefit of U.S. Provisional Application No. 61/625,270 filed Apr. 17, 2012 and titled “MID-HANGER PLATE”. U.S. Provisional Application No. 61/625,270 filed Apr. 17, 2012 titled “MID-HANGER PLATE” is hereby incorporated by reference in its entirety into the specification of this application. The following relates to the nuclear reactor arts and related arts. There is increasing interest in compact reactor designs. Benefits include: reduced likelihood and severity of abnormal events such as loss of a coolant accident (LOCA) event (both due to a reduction in vessel penetrations and the use of a smaller containment structure commensurate with the size of the compact reactor); a smaller and more readily secured nuclear reactor island (see Noel, “Nuclear Power Facility”, U.S. Pub. No. 2010/0207261 A1 published Aug. 16, 2012 which is incorporated herein by reference in its entirety); increased ability to employ nuclear power to supply smaller power grids, e.g. using a 300 MWe or smaller compact reactor, sometimes referred to as a small modular reactor (SMR); scalability as one or more SMR units can be deployed depending upon the requisite power level; and so forth. Some compact reactor designs are disclosed, for example, in Thome et al., “Integral Helical-Coil Pressurized Water Nuclear Reactor”, U.S. Pub. No. 2010/0316181 A1 published Dec. 16, 2010 which is incorporated by reference in its entirety; Malloy et al., “Compact Nuclear Reactor”, U.S. Pub. No. 2012/0076254 A1 published Mar. 29, 2012 which is incorporated by reference in its entirety. These compact reactors are of the pressurized water reactor (PWR) type in which a nuclear reactor core is immersed in primary coolant water at or near the bottom of a pressure vessel, and the primary coolant is suitably light water maintained in a subcooled liquid phase in a cylindrical pressure vessel that is mounted generally upright (that is, with its cylinder axis oriented vertically). A hollow cylindrical central riser is disposed concentrically inside the pressure vessel and (together with the core basket or shroud) defines a primary coolant circuit in which coolant flows upward through the reactor core and central riser, discharges from the top of the central riser, and reverses direction to flow downward back to below the reactor core through a downcomer annulus defined between the pressure vessel and the central riser. The nuclear core is built up from multiple fuel assemblies each comprising a bundle of fuel rods containing fissile material (typically 235U). The compact reactors disclosed in Thome et al. and Malloy et al. are integral PWR designs in which the steam generator(s) is disposed inside the pressure vessel, namely in the downcomer annulus in these designs. Integral PWR designs eliminate the external primary coolant loop carrying radioactive primary coolant. The designs disclosed in Thome et al. and Malloy et al. employ internal reactor coolant pumps (RCPs), but use of external RCPs (e.g. with a dry stator and wet rotor/impeller assembly, or with a dry stator and dry rotor coupled with a rotor via a suitable mechanical vessel penetration) is also contemplated (as is a natural circulation variant that does not employ RCPs). The designs disclosed in Thome et al. and Malloy et al. further employ internal pressurizers in which a steam bubble at the top of the pressure vessel is buffered from the remainder of the pressure vessel by a baffle plate or the like, and heaters, spargers, or so forth enable adjustment of the temperature (and hence pressure) of the steam bubble. The internal pressurizer avoids large diameter piping that would otherwise connect with an external pressurizer. In a typical PWR design, upper internals located above the reactor core include control rod assemblies with neutron-absorbing control rods that are inserted into/raised out of the reactor core by control rod drive mechanisms (CRDMs). These upper internals include control rod assemblies (CRAs) comprising neutron-absorbing control rods yoked together by a spider. Conventionally, the CRDMs employ motors mounted on tubular pressure boundary extensions extending above the pressure vessel, which are connected with the CRAs via suitable connecting rods. In this design, the complex motor stator can be outside the pressure boundary and magnetically coupled with the motor rotor disposed inside the tubular pressure boundary extension. The upper internals also include guide frames constructed as plates held together by tie rods, with passages sized to cam against and guide the translating CRA's. For compact reactor designs, it is contemplated to replace the external CRDM motors with wholly internal CRDM motors. See Stambaugh et al., “Control Rod Drive Mechanism for Nuclear Reactor”, U.S. Pub. No. 2010/0316177 A1 published Dec. 16, 2010 which is incorporated herein by reference in its entirety; and DeSantis, “Control Rod Drive Mechanism for Nuclear Reactor”, U.S. Pub. No. 2011/0222640 A1 published Sep. 15, 2011 which is incorporated herein by reference in its entirety. Advantageously, only electrical vessel penetrations are needed to power the internal CRDM motors. In some embodiments, the scram latch is hydraulically driven, so that the internal CRDM also requires hydraulic vessel penetrations, but these are of small diameter and carry primary coolant water as the hydraulic working fluid. The use of internal CRDM motors shortens the connecting rods, which reduces the overall weight, which in turn reduces the gravitational impetus for scram. To counteract this effect, some designs employ a yoke that is weighted as compared with a conventional spider, and/or may employ a weighted connecting rod. See Shargots et al., “Terminal Elements for Coupling Connecting Rods and Control Rod Assemblies for a Nuclear Reactor”, U.S. Pub. No. 2012/0051482 A1 published Mar. 1, 2012 which is incorporated herein by reference in its entirety. Another design improvement is to replace the conventional guide frames which employ spaced apart guide plates held together by tie rods with a continuous columnar guide frame that provides continuous guidance to the translating CRA's. See Shargots et al, “Support Structure for a Control Rod Assembly of a Nuclear Reactor”, U.S. Pub. No. 2012/0099691 A1 published Apr. 26, 2012 which is incorporated herein by reference in its entirety. The use of internal CRDMs and/or continuous guide frames and/or internal RCPs introduces substantial volume, weight, and complexity to the upper internals. These internals are “upper” internals in that they are located above the reactor core, and they must be removed prior to reactor refueling in order to provide access to the reactor core. In principle, some components (especially the internal RCPs) can be located below the reactor core, but this would introduce vessel penetrations below the reactor core which is undesirable since a LOCA at such low vessel penetrations can drain the primary coolant to a level below the top of the reactor core, thus exposing the fuel rods. Another option is to employ external RCPs, but this still leaves the complex internal CRDMs and guide frames. Disclosed herein are improvements that provide various benefits that will become apparent to the skilled artisan upon reading the following. In a disclosed aspect, an apparatus comprises: a pressure vessel comprising an upper vessel section and a lower vessel section; a nuclear reactor core comprising fissile material disposed the lower vessel section; and upper internals disposed in the lower vessel section above the nuclear reactor core and mounted on a suspended support assembly including a plurality of hanger plates connected by tie rods. The upper internals include at least guide frames and internal control rod drive mechanisms (CRDMs) with CRDM motors. The plurality of hanger plates includes a mid-hanger plate that is not the uppermost plate of the plurality of hanger plates and is not the lowermost plate of the plurality of hanger plates. The internal CRDMs are disposed above the mid-hanger plate, the guide frames are disposed below the mid-hanger plate, and the mid-hanger plate engages both the internal CRDMs and the guide frames. In another disclosed aspect, an apparatus comprises upper internals configured to be disposed as a unit in a nuclear reactor. The upper internals include: a suspended support assembly including a plurality of hanger plates connected by tie rods, the hanger plates including a mid-hanger plate, an upper hanger plate disposed above the mid-hanger plate, and a lower hanger plate disposed below the mid-hanger plate; guide frames mounted to the mid-hanger plate and the lower hanger plate; and internal control rod drive mechanisms (CRDMs) with CRDM motors, the internal CRDMs mounted to the mid-hanger plate and the upper hanger plate. In another disclosed aspect, an apparatus comprises: a support plate; guide frames top supported by and hanging down from the support plate; and internal control rod drive mechanisms (CRDMs) with CRDM motors, the internal CRDMs being bottom supported by the support plate. In some embodiments there is a one-to-one correspondence between internal CRDMs and guide frames, and each internal CRDM and corresponding guide frame are aligned with an opening in the support plate. With reference to FIG. 1, a small modular reactor (SMR) 1 of the of the integral pressurized water reactor (PWR) variety is shown in partial cutaway to reveal selected internal components. The illustrative PWR 1 includes a nuclear reactor core 2 disposed in a pressure vessel comprising a lower vessel portion 3 and an upper vessel portion 4. The lower and upper vessel portions 3, 4 are connected by a mid-flange 5. Specifically, a lower flange 5L at the open top of the lower vessel portion 3 connects with the bottom of the mid-flange 5, and an upper flange 5U at the open bottom of the upper vessel portion 4 connects with a top of the mid-flange 5. The reactor core 2 is disposed inside and at or near the bottom of the lower vessel portion 3, and comprises a fissile material (e.g., 235U) immersed in primary coolant water. A cylindrical central riser 6 is disposed coaxially inside the cylindrical pressure vessel and a downcomer annulus 7 is defined between the central riser 6 and the pressure vessel. The illustrative PWR 1 includes internal control rod drive mechanisms (internal CRDMs) 8 with internal motors 8m immersed in primary coolant that control insertion of control rods to control reactivity. Guide frames 9 guide the translating control rod assembly (e.g., each including a set of control rods comprising neutron absorbing material yoked together by a spider and connected via a connecting rod with the CRDM). The illustrative PWR 1 employs one or more internal steam generators 10 located inside the pressure vessel and secured to the upper vessel portion 4, but embodiments with the steam generators located outside the pressure vessel (i.e., a PWR with external steam generators) are also contemplated. The illustrative steam generator 10 is of the once-through straight-tube type with internal economizer, and is fed by a feedwater inlet 11 and deliver steam to a steam outlet 12. See Malloy et al., U.S. Pub. No. 2012/0076254 A1 published Mar. 29, 2012 which is incorporated by reference in its entirety. The illustrative PWR 1 includes an integral pressurizer 14 at the top of the upper vessel section 4 which defines an integral pressurizer volume 15; however an external pressurizer connected with the pressure vessel via suitable piping is also contemplated. The primary coolant in the illustrative PWR 1 is circulated by reactor coolant pumps (RCPs) comprising in the illustrative example external RCP motors 16 driving an impeller located in a RCP plenum 17 disposed inside the pressure vessel. With reference to FIGS. 2 and 3, a variant PWR design 1′ is shown, which differs from the PWR 1 of FIG. 1 by having a differently shaped upper vessel section 4′ and internal RCPs 16′ in place of the external pumps 16, 17 of the PWR 1. FIG. 2 shows the pressure vessel with the upper vessel section 4′ lifted off, as is done during refueling. The mid-flange 5 remains disposed on the lower flange 5L of the lower vessel 3. FIG. 3 shows an exploded view of the lower vessel section 3 and principle components contained therein, including: the nuclear reactor core 2 comprising fuel assemblies 2′ contained in a core former 20 disposed in a core basket 22. With continuing reference to FIGS. 1 and 3 and with further reference to FIGS. 4 and 5, above the reactor core assembly 2, 20, 22 are the upper internals which include a suspended support assembly 24 comprising an upper hanger plate 30, a mid-hanger plate 32, and a lower hanger plate 34 suspended by tie rods 36 from the mid-flange 5. More particularly, in the illustrative embodiment the upper ends of the tie rods 36 are secured to a riser transition section 38 that is in turn secured with the mid-flange 5. The central riser 6 disposed in the upper vessel section 4, 4′ (shown only in FIG. 1) is connected with the core basket 22 in the lower vessel section 3 by the riser cone (not shown) and riser transition section 38 to form a continuous hollow cylindrical flow separator between the columnar hot leg of the primary coolant path flowing upward and the cold leg that flows through the downcomer annulus surrounding the hot leg. The suspended support assembly 24 comprising hanger plates 30, 32, 34 interconnected by tie rods 36 provides the structural support for the CRDMs 8 and the guide frames 9 (note the CRDMs 8 and guide frames 9 are omitted in FIG. 3). The CRDMs 8 are disposed between the upper hanger plate 30 and the mid-hanger plate 32, and are either (1) top-supported in a hanging fashion from the upper hanger plate or (2) bottom-supported on the mid-hanger plate 32 (as in the illustrative embodiments described herein). Lateral support for the CRDMs 8 is provided by both plates 30, 32. (Note that in the illustrative embodiment, the CRDMs 8 actually pass through openings of the upper hanger plate 30 so that the tops of the CRDMs 8 actually extend above the upper hanger plate 30, as best seen in FIG. 1). The guide frames 9 are disposed between the mid-hanger plate 32 and the lower hanger plate 34, and are likewise either (1) top-supported in a hanging fashion from the mid-hanger plate 32 (as in the illustrative embodiments described herein) or (2) bottom-supported on the lower hanger plate. Lateral support for the guide frames 9 is provided by both plates 32, 34. One of the hanger plates, namely the mid-hanger plate 32 in the illustrative embodiments, also includes or supports a distribution feature or plate that includes mineral insulated cabling (MI cables) for delivering electrical power to the CRDM motors 8M and, in some embodiments, hydraulic lines for delivering hydraulic power to scram latches of the CRDMs 8. In the embodiment of FIGS. 2 and 3 (and as seen in FIG. 3), the internal RCPs 16′ are also integrated into the upper internals assembly 24, for example on an annular pump plate providing both separation between the suction (above) and discharge (below) sides of the RCPs 16′ and also mounting supports for the RCPs 16′. The disclosed upper internals have numerous advantages. The suspension frame 24 hanging from the mid-flange 5 is a self-contained structure that can be lifted out of the lower vessel section 3 as a unit during refueling. Therefore, the complex assembly of CRDMs 8, guide frames 9, and ancillary MI cabling (and optional hydraulic lines) does not need to be disassembled during reactor refueling. Moreover, by lifting the assembly 5, 24, 8, 9 out of the lower vessel 3 as a unit (e.g. using a crane) and moving it to a suitable work stand, maintenance can be performed on the components 5, 24, 8, 9 simultaneously with the refueling, thus enhancing efficiency and speed of the refueling. The tensile forces in the tie rods 36 naturally tend to laterally align the hanger plates 30, 32, 34 and thus the mounted CRDMs 8 and guide frames 9. The upper internals are thus a removable internal structure that is removed as a unit for reactor refueling. The upper internals basket (i.e., the suspension frame 24) is advantageously flexible to allow for movement during fit-up when lowering the upper internals into position within the reactor. Toward this end, the horizontal plates 30, 32, 34 are positioned at varying elevations and are connected to each other, and the remainder of the upper internals, via the tie rods 36. The design of the illustrative upper internals basket 24 is such that the control rod guide frames 9 are hung from the mid-hanger plate 32 (although in an alternative embodiment the guide frames are bottom-supported by the lower hanger plate). In the top-supported hanging arrangement, the guide frames 9 are laterally supported at the bottom by the lower hanger plate 34. The upper internals are aligned with the core former 20 and/or core basket 22 to ensure proper fit-up of the fuel to guide frame interface. This alignment is achieved by keying features of the lower hanger plate 34. With reference to FIGS. 6 and 7, alternative perspective views are shown of the hanger plates 30, 32, 34 connected by tie rods 36 and with the guide frames 9 installed, but omitting the CRDMs 8 so as to reveal the top surface of the mid-hanger plate 32. In the illustrative embodiment, a distribution plate 40 is disposed on top of the mid-hanger plate 32, as best seen in FIG. 6. The distribution plate 40 is a load-transferring element that transfers (but does not itself support) the weight of the bottom-supported CRDMs 8 to the mid-hanger plate 32. This is merely an illustrative example, and the distribution plate can alternatively be integral with the mid-hanger plate (e.g., comprising MI cables embedded in the mid-hanger plate) or located on or in the upper hanger plate. (Placement of the distribution plate in the lower hanger plate is also contemplated, but in that case MI cables would need to run from the distribution plate along the outsides of the guide frames to the CRDMs. As yet another option, the distribution plate can be omitted entirely in favor of discrete MI cables run individually to the CRDMs 8). With reference to FIG. 8, which shows a corner of the upper hanger plate 30 as an illustrative example, the tie rods 36 are coupled to each plate by tie rod couplings 42, which optionally incorporate a turnbuckle (i.e. length adjusting) arrangement as described elsewhere herein. Note that the ends of the tie rods connect with a hanger plate, with no hanger plate connecting at a middle of a tie rod. Thus, the upper tie rods 36 extend between the upper and mid-hanger plates 30, 32 with their upper ends terminating at tie rod couplings 42 at the upper hanger plate 30 and their lower ends terminating at tie rod couplings 42 at the mid-hanger plate 32; and similarly, the lower tie rods 36 extend between the mid-hanger plate 32 and the lower hanger plate 34 with their upper ends terminating at tie rod couplings 42 at the mid-hanger plate 32 and their lower ends terminating at tie rod couplings 42 at the lower hanger plate 34. With reference to FIGS. 9 and 10, the lower hanger plate 34 in the illustrative embodiment provides only lateral support for the guide frames 9 which are top-supported in hanging fashion from the mid-hanger plate 32. Consequentially, the lower hanger plate 34 is suitably a single plate with openings 50 that mate with the bottom ends of the guide frames (see FIG. 10). To simplify the alignment, in some embodiments guide frame bottom cards 52 (see FIG. 9) are inserted into the openings 50 and are connected with the bottom ends of the guide frames 9 by fasteners, welding, or another technique. (Alternatively, the ends of the guide frames may directly engage the openings 50 of the lower hanger plate 34). In addition to providing lateral support for each control rod guide frame 9, locking each in laterally with a honeycomb-type structure (see FIG. 10), the lower hanger plate 34 also includes alignment features 54 (see FIG. 10) that align the upper internals with the core former 20 or with the core basket 22. The illustrative alignment features are peripheral notches 54 that engage protrusions (not shown) on the core former 20; however, other alignment features can be employed (e.g., the lower hanger plate can include protrusions that mate with notches of the core former). Also seen in FIG. 10 are peripheral openings 56 in the lower hanger plate 34 into which the tie rod couples 42 of the lower hanger plate fit. The lower hanger plate 34 is suitably machined out of plate material or forging material. For example, in one contemplated embodiment the lower hanger plate 34 is machined from 304L steel plate stock. With continuing reference to FIGS. 6 and 7 and with further reference to FIG. 11, the mid-hanger plate 32 provides top support for the guide frames 9 and bottom support for the CRDMs 8. The mid-hanger plate 32 acts as a load distributing plate taking the combined weight of the CRDMs 8 and the guide frames 9 and transferring that weight out to the tie rods 36 on the periphery of the upper internals basket 24. In the illustrative embodiment, the power distribution plate 40 is also bottom supported. Like the lower hanger plate 34, the mid-hanger plate 32 includes openings 60. The purpose of the openings 60 is to enable the connecting rod, translating screw, or other coupling mechanism to connect each CRDM 8 with the control rod assembly driven by the CRDM. To facilitate hanging the guide frames 9 off the bottom of the mid-hanger plate 32, an egg crate-type structure made of orthogonally intersecting elements 61 is provided for increased strength and reduced deflection due to large loads. With reference to FIGS. 12 and 13, the mid-hanger plate 32 can be manufactured in various ways. In one approach (FIG. 12), a forging machining process is employed to machine the mid-hanger plate 32 out of a 304L steel forged plate 62. The machining forms the openings 60 and the intersecting elements 61. In another approach (FIG. 13), a machined plate 64 and the intersecting elements 61 are manufactured as separate components, and the intersecting elements 61 are interlocked using mating slits formed into the intersecting elements 61 and welded to each other and to the machined plate 64 to form the mid-hanger plate 32. As previously noted, the illustrative bottom-supported distribution plate 40 can alternatively be integrally formed into the mid-hanger plate. With reference to FIG. 14, in an alternative embodiment the guide frames 9 are bottom supported by an alternative lower hanger plate 34′, and are laterally aligned at top by an alternative mid-hanger plate 32′. In this case the alternative lower hanger plate 34′ may have the same form and construction as the main embodiment mid-hanger plate 32 of FIGS. 11-13 (but with suitable alignment features to align with the core former and/or core basket, not shown in FIG. 14), and the alternative mid-hanger plate 32′ can have the same form and construction as the main embodiment lower hanger plate 34 of FIG. 10 (but without said alignment features). If the CRDMs remain bottom supported, then the alternative mid-hanger plate 32′ should be made sufficiently thick (or otherwise sufficiently strong) to support the weight of the CRDMs. As another variant, the alternative mid-hanger plate 32′ can be made too thin to directly support the CRDMs, and an additional thicker upper plate added to support the weight of the CRDMs. In this case the thicker plate would be the one connected with the tie rods to support the CRDMs. In the illustrative embodiments, the guide frames 9 are continuous columnar guide frames 9 that provide continuous guidance to the translating control rod assemblies. See Shargots et al, “Support Structure for a Control Rod Assembly of a Nuclear Reactor”, U.S. Pub. No. 2012/0099691 A1 published Apr. 26, 2012 which is incorporated herein by reference in its entirety. However, the described suspended frame 24 operates equally well to support more conventional guide frames comprising discrete plates held together by tie rods. Indeed, the main illustrative approach in which the guide frames are top-supported in hanging fashion from the mid-hanger plate 32 is particularly well-suited to supporting conventional guide frames, as the hanging arrangement tends to self-align the guide frame plates. With reference to FIG. 15, an illustrative embodiment of the upper hanger plate 30 is shown. Like the other hanger plates 32, 34, the upper hanger plate 30 includes openings 70, in this case serving as passages through which the upper ends of the CRDMs 8 pass. The inner periphery of each opening 70 serves as a cam to laterally support and align the upper end of the CRDM 8. The upper hanger plate 30 can also suitably be made by machining from either plate material or forging material, e.g. a 304L steel plate stock or forging. With reference to FIGS. 16-18, the tie bar (alternatively “tie rod”) couplings 42 are further described. FIG. 16 shows the suspended frame 24 including the upper, mid-, and lower hanger plates 30, 32, 34 held together by tie rods 36. For clarity, the tie bars are denoted in FIG. 16 as upper tie bars 361 and lower tie bars 362, and the various levels of tie bar couples are denoted as upper tie bar couples 421, middle tie bar couples 422, and lower tie bar couples 423. At the upper end, short tie rods (i.e. tie rod bosses) 36B have upper ends welded to the riser transition 38 and have lower ends threaded into the tops of upper tie bar couplings 421. The upper tie bars 361 have their upper ends threaded into the bottoms of upper tie bar couplings 421 and have their lower ends threaded into the tops of middle tie bar couplings 422. The lower tie bars 362 have their upper ends threaded into the bottoms of middle tie bar couplings 422 and have their lower ends threaded into the tops of lower tie bar couplings 423. FIGS. 17 and 18 show perspective and sectional perspective views, respectively, of the middle tie bar coupling 422. As best seen in FIG. 18, the tie rod coupling 422 has a turnbuckle (i.e. length adjusting) configuration including outer sleeves 81, 82 having threaded inner diameters that engage (1) the threaded outsides of the ends of the respective mating tie rods 361, 362, and (2) the threaded outsides of a plate thread feature 84. Thus, by rotating the outer sleeve 81 the position of tie rod 361 respective to the mid-hanger plate 32 can be adjusted; and similarly, by rotating the outer sleeve 82 the position of tie rod 362 respective to the mid-hanger plate 32 can be adjusted. (Note that the plate thread feature 84 can be a single element passing through the mid-hanger plate 32, or alternatively can be upper and lower elements extending above and below the mid-hanger plate 32, respectively). The tie bar coupling 421 is the same as tie bar coupling 422 except that the upper outer sleeve 81 suitably engages the tie rod boss 36B; while, the tie bar coupling 42 is the same as tie bar coupling 422 but omits the lower half (i.e. lower outer sleeve 82 and the corresponding portion of the plate thread feature 84), since there is no tie rod “below” for the tie bar coupling 423 to engage. Said another way, the tie rod coupling portions 81, 82 can be threaded on their inner diameter with threads matching that of the outer diameter of the tie rods 36 and on the threading feature 84 of any of the plates 30, 32, 34 or riser transition 38. This allows the coupling 42 to be threaded onto the tie rod 36 and onto the threading feature 84 of any other component. The advantages to a coupling such as this is that a very accurate elevation can be held with each of the above mentioned components 30, 32, 34, 38 within the upper internals, and that each of the above components can hold a very accurate parallelism with one another. Essentially, the couplings allow for very fine adjustments during the final assembly process. They also allow for a quick and easy assembly process. Another advantage to the couplings 42 is that they allow for the upper internals to be separated at the coupling joints fairly easily for field servicing or decommissioning of the nuclear power plant. In an alternative tie rod coupling approach, it is contemplated for the tie rods to be directly welded to any of the plates or riser transition, in which case the tie rod couplings 42 would be suitably omitted. However, this approach makes it difficult to keep the tie rod perpendicular to the plates making assembly of the upper internals more difficult. It also makes breaking the upper internals down in the field more difficult. With reference to FIG. 19, the riser transition 38 is shown in perspective view. The riser transition assembly 38 performs several functions. The riser transition 38 provides load transfer from the tie rods 36 of the upper internals basket 24 to the mid-flange 5 of the reactor pressure vessel. Toward this end, the riser transition 38 includes gussets 90 by which the riser transition 38 is welded to the mid-flange 5. (See also FIGS. 4 and 5 showing the riser transition 38 with gussets 90 welded to the mid-flange 5). One or more of these gussets 90 may include a shop lifting lug 91 or other fastening point to facilitate transport, for example when the upper internals are lifted out during refueling. The load transfer from the tie rods 36 to the mid-flange 5 is mostly vertical loading due to the overall weight of the upper internals. However, there is also some radial differential of thermal expansion between the riser transition gussets 90 and the mid-flange 5, and the riser transition 38 has to also absorb these thermal loads. As already mentioned, the riser cone and riser transition 38 also acts (in conjunction with the central riser 6 and core basket 22) as the flow divider between the hot leg and cold leg of the primary coolant loop. Still further, the riser transition 38 also houses or includes an annular hydraulic collection header 92 for supplying hydraulic power via vertical hydraulic lines 94 to the CRDMs (in the case of embodiments employing hydraulically driven scram mechanisms). The riser transition 38 also has an annular interface feature 96 for fit-up with the riser cone or other connection with the central riser 6, and feature cuts 98 to allow the passing of the CRDM electrical MI cable. With brief returning reference to FIGS. 4 and 5, the gussets 90 are suitably welded to the mid-flange 5 at one end and welded to the main body portion of the riser transition assembly 38 at the other end. The riser transition 38 is suitably made of 304L steel, in some embodiments, e.g. by machining from a ring forging. With reference to FIG. 20, an illustrative gusset 90 is shown, having a first end 100 that is welded to the mid-flange 5 and a second end 102 that is welded to the riser transition 38 as already described. The gusset 90 includes horizontal cantilevered portion 104, and a tensile-strained portion 106 that angles generally downward, but optionally with an angle A indicated in FIG. 20. The horizontal cantilevered portion 104 has a thickness dcant that is relatively greater than a thickness dG of the tensile-strained portion 106. The thicker cantilevered portion 104 handles the vertical loading component, while the tensile-strained portion 106 allows the gusset 90 to deflect in the lateral direction to absorb lateral loading due to thermal expansion. The angle A of the tensile-strained portion 106 provides for riser cone lead-in. The end 102 of the gusset 90 that is welded to the riser transition 38 includes an upper ledge 108 that serves as a riser cone interface. In the illustrative embodiments, the CRDMs 8 are bottom supported from the mid-hanger plate 32, and the tops of the CRDMs 8 are supported by the upper hanger plate 30, which serves as the lateral support for each CRDM, locking each in laterally with a honeycomb type structure (see FIG. 15). Even with this support structure, however, the CRDM 8 should be protected during an Operating Basis Earthquake (OBE) or other event that may cause mechanical agitation. To achieve this, it is desired to support the upper end of the CRDM to prevent excessive lateral motion and consequently excessive loads during an OBE. It is disclosed to employ a restraining device which still allows for ease of maintenance during an outage. Using spring blocks integrated into the CRDM 8 satisfies both of these requirements, as well as providing compliance that accommodates any differential thermal expansion. Integrating compliance features into support straps of the CRDM 8 allows the CRDM's to be removed while still maintaining lateral support. As the CRDM is lowered into its mounting location the compliant features come into contact with the upper hanger plate 30. The compliance allows them to maintain contact with the upper hanger plate yet allow for misalignment between the CRDM standoff mounting point and the upper hanger plate. Their engagement into the upper hanger plate 30 allows them to be of sufficient height vertically from the mounting base of the CRDMs to minimize the loads experienced at the base in an OBE event. Having no feature that extends below the upper hanger plate allows the CRDM to be removed from the top for service. With reference to FIGS. 21 and 22, an upper end of a CRDM 8 includes a hydraulic line 110 delivering hydraulic power to a scram mechanism. Straps 112, 114 secure the hydraulic line 110 to the CRDM 8. The strap 114 is modified to include compliance features 116. As seen in FIG. 22, the compliance features 116 comprise angled spring blocks that wedges into the opening 70 of the upper hanger plate 30 when the CRDM 8 is fully inserted. It will be appreciated that such compliance features 116 can be incorporated into straps retaining other elements, such as electrical cables (e.g. MI cables). The illustrative compliance features 116 can be constructed as angled leaf springs cut into the (modified) strap 114. Alternatively, such leaf springs can be additional elements welded onto angled ends of the strap 114. By including such springs on straps 114 on opposite sides of the CRDM 8, four contact points are provided to secure the CRDM against lateral motion in any direction. The wedged support provided by the straps 114 also leave substantial room for coolant flow through the opening 70 in the upper hanger plate 30. The disclosed embodiments are merely illustrative examples, and numerous variants are contemplated. For example, the suspended frame of the upper internals can include more than three plates, e.g. the power distribution plate could be a separate fourth plate. In another variant, the mid-hanger plate 32 could be separated into two separate hanger plates—an upper mid-hanger plate bottom-supporting the CRDMs, and a lower mid-hanger plate from which the guide frames are suspended. In such a case, the two mid-hanger plates would need to be aligned by suitable alignment features to ensure relative alignment between the CRDMs and the guide frames. The use of at least three hanger plates is advantageous because it provides both top and bottom lateral support for both the CRDMs and the guide frames. However, it is contemplated to employ only two hanger plates if, for example, the bottom support of the CRDMs is sufficient to prevent lateral movement of the CRDMs. In the illustrative embodiments, the suspended support assembly 24 is suspended from the mid-flange 5 via the riser transition 38. However, other anchor arrangements are contemplated. For example, the suspended support assembly could be suspended directly from the mid-flange, with the riser transition being an insert secured to the gussets. The mid-flange 5 could also be omitted. One way to implement such a variant is to include a ledge in the lower vessel on which a support ring sits, and the suspended support assembly is then suspended from the support ring. With the mid-flange 5 omitted, the upper and lower flanges 5U, 5L of the upper and lower vessel sections can suitably connect directly (i.e., without an intervening mid-flange). Instead of lifting the upper internals out by the mid-flange 5, the upper internals would be lifted out by the support ring. In the embodiment of FIGS. 2 and 3, the internal RCPs 16′ are incorporated into the upper internals and are lifted out with the upper internals. Other configurations are also contemplated—for example, internal RCPs could be mounted in the upper vessel and removed with the upper vessel. The preferred embodiments have been illustrated and described. Obviously, modifications and alterations will occur to others upon reading and understanding the preceding detailed description. It is intended that the invention be construed as including all such modifications and alterations insofar as they come within the scope of the appended claims or the equivalents thereof.
040452867	claims	1. A molten-salt reactor provided within a common vessel containing the reactor core and a neutron-moderating mass pierced by passages for the circulation of the molten fuel salt with at least one primary heat exchanger which is located as close as possible to the reactor core and through which the hot fuel salt passes immediately after discharge from said core and with pumps for circulating the cold molten fuel salt which is discharged from the heat exchangers and returned into the reactor core, wherein the free spaces defined within the reactor vessel exterior of and between the core, the heat exchangers and the pumps are filled with expanded graphite which is compatible with the molten fuel salt except for the connecting passages in which the fuel salt is circulated, wherein the reactor core is placed within the central portion of an open vessel having a vertical axis and is surrounded by a lateral reflector which defines an annular region with the internal vessel wall, the pumps for the circulation of molten fuel salt and the heat exchangers being placed within said annular region, said pumps and said heat exchangers being disposed at intervals around said reactor core. 2. A molten-salt reactor according to claim 1, wherein the pumps and the heat exchangers are suspended within the annular space beneath a horizontal vault roof extending above the reactor vessel, said vault roof being provided with a central access opening placed opposite to the reactor core and closed by a removable shield plug. 3. A molten-salt reactor according to claim 1, wherein each circulating pump is mounted directly beneath a heat exchanger within the annular space so as to constitute a pump-exchanger unit, the passages providing a connection with the reactor core being constituted by ducts extending radially from the axis of the reactor vessel and placed in the top and bottom portions of said reactor core. 4. A molten-salt reactor according to claim 1, wherein the reactor vessel wall is provided with an internal heat insulation and an external cooling circuit which is capable of producing partial solidification of the fuel salt in contact with said wall so as to limit the effects of corrosion by the circulating molten salt.
	abstract	The radiation image conversion panel includes a substrate and a phosphor layer of columnar crystals formed on the substrate by vapor-phase deposition, with a column diameter distribution of the columnar crystals having two or more peaks. The process for producing a radiation image conversion panel prepares a substrate on which two or more types of projections different in diameter are formed and satisfies Expression “0.4R≦r≦0.8R” where R is a diameter of a largest projection and r is a diameter of any one of the remainder in the two or more types of projections, thereby making a surface of the substrate uneven and forms a phosphor layer on the uneven surface of the substrate by vapor-phase deposition.
	abstract	An X-ray mask structure and X-ray exposure method using the same are disclosed, wherein the mask has an X-ray absorptive material pattern, a supporting film for supporting the pattern, and a holding frame for holding the supporting film, wherein a suction port is arranged to be communicated with an external gas drawing system, and wherein a supply port is provided so that a gas can be supplied therethrough, for prevention of dust adhesion to the mask.
	claims	1. A laser processing device including a controller for controlling a laser generator for generating laser light adapted to process a processing target, and a head with a scanning optical system for scanning the laser light with respect to the processing target and a housing for accommodating the scanning optical system; the laser processing device comprising:a camera for photographing the processing target, the camera being removably attached to the housing and having a light receiving axis branched from an emission axis of the laser light;a camera cover being removably attached to the housing to cover the camera;a cover detection section for detecting detachment of the camera cover from the housing; anda laser output control section for prohibiting emission of the laser light to the processing target based on the detection result of the cover detection section. 2. The laser processing device according to claim 1, further comprising a shutter for blocking a light receiving path of the camera in an openable/closable manner and blocking the light receiving path upon output of the laser light. 3. The laser processing device according to claim 1, whereinthe camera cover includes a projection; andthe cover detection section includes a mechanical contact that shifts to a conducting state when the camera cover is attached to the housing and the projection is brought into contact, and shifts to a blocking state when the camera cover is detached from the housing. 4. The laser processing device according to claim 1, whereinthe laser generator includes an excitation light source for generating excitation light and a laser oscillator for generating the laser light based on the excitation light;an excitation light source power supply for supplying power to the excitation light source using a commercial power supply is further arranged; andthe laser output control section controls the power supply with respect to the excitation light source power supply based on the detection result of the cover detection section. 5. The laser processing device according to claim 4, whereinthe excitation light source includes a light emitting element and a drive circuit for supplying a drive current to the light emitting element; andthe laser output control section instructs a current value of the drive current with respect to the drive circuit based on the detection result of the cover detection section. 6. The laser processing device according to claim 1, further comprising a notifying section for notifying that the camera cover is detached from the housing based on the detection result of the cover detection section. 7. The laser processing device according to claim 1, further comprising a camera mount for arranging the camera on an exterior of the housing, the camera mount including an offset adjustment mechanism for making a light receiving axis of the camera side substantially coincide with a light receiving axis of the housing side and an angle adjustment mechanism for adjusting an attachment angle of the camera with the light receiving axis as a center. 8. The laser processing device according to claim 1, further comprising:a telecentric lens for making an emission angle of the laser light constant irrespective of an incident angle of the laser light, the telecentric lens being arranged on the processing target side than the scanning optical system; andan optical splitter for branching a light receiving path of the camera from the emission path of the laser light, the optical splitter being arranged on the laser generator side than the scanning optical system.
056082240	abstract	A target changer for use with an accelerator for changing targets remotely. The target changer includes a beam tube, an end of which is secured to a ring collimator assembly, a carousel barrel which defines a plurality of ports for receiving targets, a carousel hub for permitting the rotation of the carousel barrel and alignment with the ring collimator assembly and beamline, and a motor for controlling the rotation of the carousel barrel.
	summary
046559957	claims	1. A nuclear fuel assembly comprising: (a) a flow channel; (b) a lower nozzle assembly structurally attached to said flow channel to form therewith an external envelope; (c) an invertible fuel bundle adapted to be inserted into said envelope, said fuel bundle comprising a plurality of elongated fuel rods held in a spaced lateral array between top and bottom tie plates, each of said top and bottom tie plates being substantially identical and having means for supporting said fuel bundle within said envelope in either of two mututally inverted vertical orientations whereby the orientation of the fuel bundle in a flow channel may be reversed during burn-up operation. (a) inserting a nuclear fuel bundle having substantially identical top and bottom tie plates into said external envelope with one of said top and bottom tie plates engaging said bottom nozzle assembly to support said fuel bundle thereon; (b) operating said reactor for a period of time; (c) withdrawing said fuel bundle from said envelope; (d) inverting said fuel bundle; (e) reinserting said inverted fuel bundle into said envelope with the other of said top and bottom tie plates engaging said bottom nozzle to support said fuel assembly thereon; and (f) continuing to operate said reactor with said inverted fuel bundle. (a) providing fission gas plenums in each fuel rod of said fuel bundle; (b) disposing some of said fission gas plenums at the top of said fuel bundle and a remainer of said fission gas plenums at the bottom of said fuel bundle. (a) orienting alternate fission gas plenums at the top of the fuel bundle; and (b) orienting intervening alternate fission gas plenums at the bottom of the fuel bundle. (a) inserting the self-supporting nuclear fuel bundle having substantially identical top and bottom tie plates into said fuel bundle envelope with one of said top and bottom tie plates engaging said bottom nozzle assembly to support said fuel bundle thereron; (b) operating said reactor for a period of time to partially burn said fuel whereby fuel positioned at the lower portion of said fuel bundle is burned to a larger extent than fuel positioned at an upper portion of said fuel bundle; (c) withdrawing said fuel bundle from said envelope; (d) inverting the top and bottom of said fuel bundle; (e) reinserting said inverted fuel bundle into said envelope whereby the fuel burned to a larger extent is positioned at the upper portion of the fuel bundle; (f) continuing to operate said reactor with said inverted fuel bundle whereby the fuel in said fuel bundle is more evenly burned. 2. The fuel assembly of claim 1 wherein said fuel rods comprise a tubular cladding containing nuclear fuel, each of said fuel rods having a fission gas plenum disposed within said cladding to accommodate fission gases released during operation of said fuel assembly, wherein some of said fission gas plenums are disposed adjacent said top tie plate and the remainder of said fission gas plenums are disposed adjacent said bottom tie plate. 3. The fuel assembly of claim 2 wherein fission gas plenums of alternate fuel rods are disposed adjacent said top tie plate and intervening alternate fuel rod fission gas plenums are disposed adjacent said bottom tie plate. 4. The fuel assembly of claim 1 wherein said means for supporting said fuel bundle comprise a plurality of tie plate legs extending from said top and bottom tie plates, said tie plate legs having means for engaging said lower nozzle assembly and for supporting said fuel bundle thereon. 5. The fuel assembly of claim 2 wherein said means for supporting said fuel bundle comprises a plurality of tie plate legs extending from said top and bottom tie plates, said tie plate legs having means for engaging said lower nozzle assembly and for supporting said fuel bundle. 6. The fuel assembly of claim 4 wherein said tie plate legs have lifting slots formed integrally therein. 7. The fuel assembly of claim 5 wherein said tie plate legs have lifting slots formed integrally therein. 8. The fuel assembly of claim 4 further comprising an orifice plate assembly disposed in said lower nozzle assembly and having means for engaging said tie plate legs and means for engaging said lower nozzle assembly. 9. The fuel assembly of claim 5 further comprising an orifice plate assembly disposed in said lower nozzle assembly and having means for engaging said tie plate legs and means for engaging said lower nozzle assembly. 10. The fuel assembly of claim 2 wherein each of said fuel rods further has an end plug at each end thereof, each of said end plugs having an extension extending from said cladding, said top and bottom tie plates each having a plurality of apertures therein for slidably accepting said extensions. 11. The fuel assembly of claim 2 wherein said cladding is provided with a crimp in the region of said fission gas plenums to prevent said nuclear fuel from occupying said fission gas plenums. 12. A method of operating a nuclear reactor having a core containing a fuel bundle external envelope comprising a flow channel and a bottom nozzle assembly, said method comprising the steps of: 13. The method of claim 12 further comprising the steps of: 14. The method of claim 13 wherein the step of disposing further comprises: 15. A method of operating a BWR nuclear reactor having a core containing a fuel bundle envelope comprising a flow channel and bottom nozzle assembly, said method being operable to enhance burn up of nuclear fuel and to reduce fuel cycle operating costs and comprising the steps of:
047117567	abstract	In a nuclear reactor including a core, a plurality of control rods, a support supporting the control rods and movable for displacing the control rods in their longitudinal direction between a first end position in which the control rods are fully inserted into the core and a second end position in which the control rods are retracted from the core, and guide elements contacting discrete regions of the outer surface of each control rod at least when the control rods are in the vicinity of the second end position, the control rods being longitudinally movable relative to the guide elements to thereby cause the outer surface of the control rods to experience wear as a result of sliding contact with the guide elements, there is provided a displacement device operatively coupled to the control rods for periodically rotating the control rods in order to change the locations on the outer surfaces of the control rods at which the control rods are contacted by the guide elements.
051951134	abstract	An X-ray exposure apparatus comprises an X-ray beam generating source and an exposure apparatus body including an alignment optical system. The exposure apparatus body is set independently from the X-ray beam generating source and the exposure apparatus body is located to be swingable wtih respect to the base such as a floor, through the elastic member. Thereafter, the exposure apparatus body is raised by a raising mechanism, for example, comprising a plurality of vertically expandable air cushions, in a floating manner. A height position and an inclination of the exposure apparatus body are adjusted by controlling the raising mechanism individually so that the axis of the X-ray beam generated from the X-ray source substantially coincides with the optical axis as measurement reference of the alignment optical system. The exposure apparatus body is then secured to this adjusted position by means of the securing mechanism comprising adjusting bolts and clamping bolts, for example.
	claims	1. A multileaf collimator, comprising:a leaf mounted displaceably in an adjusting linear direction that contributes to establish a contour of a beam path;a linear drive assigned to the leaf that displaces the leaf in the adjusting linear direction;a piezoelectric actuator arranged on the linear drive that moves the linear drive, wherein the piezoelectric actuator comprises a piezoelectric element and a transducer coupled thereto; anda control device that drives the piezoelectric actuator in accordance with an asymmetrically-changing voltage, the control device being operable to drive the piezoelectric element with a high speed, wherein essentially no friction is transmitted to the leaf, and being operable to drive the piezoelectric element with a low speed, wherein a frictional force is transmitted to the leaf as a driving force, wherein the asymmetrically-changing voltage is configured so that during a first state of the changing voltage the transducer overcomes a level of static frictional force between the transducer and the leaf, and, in response to overcoming said level of static frictional force, the transducer slides relative to the leaf to travel an interval in a direction opposite to the adjusting direction, wherein the asymmetrically-changing voltage is further configured so that during a second state of the changing voltage the transducer is within the level of static frictional force between the transducer and the leaf, and, in response to being within said level of static frictional force, the transducer is affixed to the leaf by way of said static frictional force to drive the leaf so that the leaf travels the interval in the adjusting direction. 2. The multileaf collimator as claimed in claim 1, wherein the transducer moves slower in a direction of motion than in an opposite direction of motion. 3. The multileaf collimator as claimed in claim 1, wherein the piezoelectric actuator is assigned to a narrow side or a flat side of the leaf. 4. The multileaf collimator as claimed in claim 1, wherein the leaf is moved by a plurality of piezoelectric actuators. 5. The multileaf collimator as claimed in claim 4, wherein the control device successively drives the piezoelectric actuators. 6. The multileaf collimator as claimed in claim 4, wherein at least two piezoelectric actuators are assigned to a narrow side or a flat side of the leaf. 7. The multileaf collimator as claimed in claim 1, wherein the multileaf collimator is used for a radiation therapy device. 8. A radiation therapy device, comprising:a retaining device; anda multileaf collimator attached on the retaining device, wherein the multileaf collimator comprises:a leaf mounted displaceably in an adjusting linear direction that contributes to establish a contour of a beam path,a linear drive assigned to the leaf that displaces the leaf in the adjusting linear direction,a piezoelectric actuator arranged on the linear drive that moves the linear drive, wherein the piezoelectric actuator comprises a piezoelectric element and a transducer coupled thereto; anda control device that drives the piezoelectric actuator in accordance with an asymmetrically-changing voltage, the control device being operable to drive the piezoelectric element with a high speed, wherein essentially no friction is transmitted to the leaf, and being operable to drive the piezoelectric element with a low speed, wherein a frictional force is transmitted to the leaf as a driving force, wherein the asymmetrically-changing voltage is configured so that during a first state of the changing voltage the transducer overcomes a level of static frictional force between the transducer and the leaf, and, in response to overcoming said level of static frictional force, the transducer slides relative to the leaf to travel an interval in a direction opposite to the adjusting direction, wherein the asymmetrically-changing voltage is further configured so that during a second state of the changing voltage the transducer is within the level of static frictional force between the transducer and the leaf, and, in response to being within said level of static frictional force, the transducer is affixed to the leaf by way of said static frictional force to drive the leaf so that the leaf travels the interval in the adjusting direction. 9. The device as claimed in claim 8, wherein the transducer moves slower in a direction of motion than in an opposite direction of motion. 10. The device as claimed in claim 8, wherein the piezoelectric actuator is assigned to a narrow side or a flat side of the leaf. 11. The device as claimed in claim 8, wherein the leaf is moved by a plurality of piezoelectric actuators. 12. The device as claimed in claim 11, wherein the control device successively drives the piezoelectric actuators. 13. The device as claimed in claim 11, wherein at least two piezoelectric actuators are assigned to a narrow side or a flat side of the leaf.
	description	This application is based upon and claims the benefit of priority from prior Japanese Patent Applications No. 2004-295250, filed on Oct. 7, 2004; and No. 2004-351724, filed on Dec. 3, 2004, the entire contents of both of which are incorporated herein by reference. 1. Field of the Invention The present invention relates to a Single Photon Emission CT (SPECT) apparatus. 2. Description of the Related Art Nuclear medicine is medicine for dosing a drug marked by radioactive isotopes (hereinafter, “RIs”) in a patient and imaging an internal RI distribution of the RIs to perform diagnosis. In a nuclear medicine diagnostic apparatus, an apparatus for imaging a three-dimensional distribution of the internal RIs is a Single Photon Emission CT (hereinafter, SPECT) apparatus. In the SPECT apparatus, filter processing is often applied to collected data of two-dimensional projection distribution to reduce a noise component and correct spatial resolution. It is effective to carry out the filter processing to projection data before reconfiguration. The filter processing is described in the following document: Edholm P E, Lewitt R M and Lindholm B: Novel properties of the Fourier decomposition of the sinogram. International Workshop on Physics and Engineering of Computerized Multidimensional Imaging and Processing. Proc of SPIE, 671, 8-18, 1986. One of causes of a fall in spatial resolution is an incidence width of a gamma ray (see FIG. 4). Incidence widths w1 and w2 change according to distances d1 and d2 between a radiation source and a detector. Spatial resolution of a radiation source S2 farther from a detector 3 than a radiation source S1 is lower than that of a radiation source S1. However, at a stage of the two-dimensional projection distribution data before reconfiguration, since the distances d1 and d2 cannot be separated, it is impossible to effectively correct the spatial resolution. It is an object of the invention to effectively correct a fall in spatial resolution in a SPECT apparatus. According to an aspect of the present invention, there is provided a SPECT apparatus including: a two-dimensional detector that detects radiations from RIs in a patient via a collimator; a correction processing unit that corrects plural two-dimensional projection distributions with different projection angles, which are detected by the detector, on a three-dimensional frequency space according to plural correction functions corresponding to plural distances, respectively; and a reconfiguring unit that reconfigures a three-dimensional RI distribution from the plural two-dimensional projection distributions corrected. Additional objects and advantages of the invention will be set forth in the description which follows, and in part will be obvious from the description, or may be learned by practice of the invention. The objects and advantages of the invention may be realized and obtained by means of the instrumentalities and combinations particularly pointed out herein after. A Single Photon Emission CT (SPECT) apparatus according to an embodiment of the invention will be hereinafter explained with reference to the accompanying drawings. Note that the invention is not limited to the SPECT apparatus. It is possible to provide the invention as a method for data processing including data correction and image reconfiguration processing in the SPECT apparatus and a program for causing a computer to realize the method. As shown in FIGS. 1 and 2, the SPECT apparatus in this embodiment includes a gamma ray detector apparatus 1 and a gantry 5 for supporting the detector apparatus 1 to freely rotate around a rotation axis (a Z axis) substantially coinciding with an axis of a patient P mounted on a top board 21 of a bed 20. A rotation coordinate system rotating around the Z axis is defined. In the rotation coordinate system, a direction axis perpendicular to a detection surface of a detector 2 is set as a Y axis and a channel direction axis of the detector 2 is set as an X axis. The detector apparatus 1 includes the two-dimensional detector 2 and a parallel hole collimator 3. The detector 2 performs two-dimensional position detection using a two-dimensional (plane) scintillator and plural photomultiplier tubes. Nal (Tl) or the like is used for the scintillator. The parallel hole collimator 3 is constituted by a lead plate with plural holes drilled in parallel in order to limit an angle of incidence of gamma rays reaching the detector 2 from radiation sources (RIs). The gantry 5 is controlled by a gantry control unit 6 to rotate the detector apparatus 1. In performing imaging, an imaging control unit 7 controls the gantry control unit 6 such that the detector apparatus 1 intermittently (or continuously) rotates at a fixed period around the patient P. A collecting unit 4 reads out a signal charge from the detector 2 and digitizes the signal charge. In addition, the collecting unit 4 discriminates events (gamma ray incidence events) in an energy window corresponding to dosed RIs and aggregates, for each stop period of the detector apparatus 1, the number of events put in the energy window for each incidence position of a gamma ray. As a result of aggregation of each cell, the number of RIs is accumulated substantially along the Y axis. As a result of aggregation, a two-dimensional spatial distribution (a two-dimensional projection distribution) of the number of RIs, which is a distribution obtained by projecting a three-dimensional RI distribution on a detector surface substantially along the Y axis, is acquired for each projection angle. Data of plural two-dimensional projection distributions with different projection angles is stored in an image storing unit 9 in association with an angle (a projection angle) of the detector 2. The imaging control unit 7 and the image storing unit 9 constitute a console box together with an input unit 11 such as a keyboard and a mouse, a control unit 8, and an image processing unit 12. The image processing unit 12 has a function of reconfiguring a three-dimensional RI distribution from plural two-dimensional projection distributions with different projection angles and correcting, in order to reduce a fall in spatial resolution according to distances between RIs and the detector 2, the plural two-dimensional projection distribution with different projection angles detected by the detector 2 according to plural correction functions corresponding to the distances on a three-dimensional frequency space. Therefore, the image processing unit 12 includes a three-dimensional filter storing unit 13, a sinogram transforming unit 14, a filter processing unit 15, a Fourier transformation unit 16, an inverse Fourier transformation unit 17, and a reconfiguration processing unit 18. The three-dimensional filter storing unit 13 stores data of plural filter functions (correction functions) corresponding to the plural distances, respectively. The sinogram transforming unit 14 transforms plural two-dimensional projection distributions (XZ surface distributions) with different projection angles into a three-dimensional projection distribution (a three-dimensional sinogram) represented by a three-dimensional actual space formed by a projection angle axis (a φ axis), a slice axis (a Z axis), and a channel axis (an X axis). The Fourier transformation unit 16 transforms the three-dimensional sinogram into a representation (γ, n, w) in a frequency space from a representation (X, φ, Z) in the actual space. γ is a frequency equivalent to a channel direction X in the actual space, n is a frequency equivalent to a projection angle φ in the actual space, w is a frequency equivalent to a slice axis direction Z in the actual space. The filter processing unit 15 corrects spatial resolution by using the plural filter functions (correction functions) stored in the three-dimensional filter storing unit 13 properly and convoluting the filer functions with respect to the three-dimensional sinogram represented in the frequency space. As described above, a degree of a fall in spatial resolution changes according to the radiation source to detector distances d. Since the distances d cannot be separated on the actual space, the fall in spatial resolution cannot be corrected effectively. However, it is possible to separate the distances d and correct the fall in spatial resolution by shifting the three-dimensional sinogram to the frequency space. The inverse Fourier transformation unit 17 transforms (returns) the three-dimensional sinogram subjected to the correction of spatial resolution into a representation in the actual space from the representation in the frequency space. The reconfiguration processing unit 18 reconfigures a three-dimensional RI distribution from the three-dimensional sinogram returned to the representation in the actual space subjected to the correction of spatial resolution. A principle of occurrence of the fall in spatial resolution is shown in FIG. 4. RIs dosed to the patient P gather in target regions. Positions of the RIs are assumed to be S1 and S2. The radiation source position S1 is apart from the detection surface of the detector 2 by a distance d1. The radiation source position S2 is apart from the detection surface of the detector 2 by a distance d2 larger than the distance d1 of the radiation source position S1. Each hole of the collimator 3 has an opening width WA. Therefore, a gamma ray radiated from the radiation source position S1 reaches the detection surface of the detector 2 in a range of a width W1. A gamma ray radiated from the radiation source position S2 is made incident on the detection surface of the detector 2 in a range of a width W2 larger than the width W1. Spatial resolution increases but sensitivity falls as an incidence width is smaller. Conversely, sensitivity increases but spatial resolution falls as an incidence width is larger. The opening width WA is designed based on sensitivity and spatial resolution that are in a trade-off relation. Therefore, it is inevitable to give a certain degree of the opening width WA to the collimator 3. As a method of correcting such a fall in spatial resolution due to an opening width of the collimator 3 according to filter processing, a Frequency Distance Relation (FDR) method is adopted. First, a principle of this FDR method is explained in an example of a two-dimensional space. Assuming that X is a detection position and φ is an angle (a projection angle) of the detector 2, when two-dimensional Fourier transformation is applied to two axis of X and φ, G(γ,n) shown in FIG. 5C is obtained with respect to a sinogram g(X,φ) shown in FIG. 5B that is obtained by expanding a one-dimensional projection distribution concerning a detection position X shown in FIG. 5A along an axis of a projection angle φ. γ represents a frequency component of a detection position and n represents a frequency component of a detector (projection) angle. In other words, plural one-dimensional projection distributions with different projection angles are transformed from a representation on the actual space into a representation on the frequency space. Note that a projection distribution of RIs represented on the frequency space is referred to as an FDR image. On the frequency space, data with the same distances d between a radiation source and a detector are present on a straight line Y=−n/γ in an FDR image G(γ,n). Y represents a distance between a line parallel to the detector 2 and including a rotation center and the radiation source. A fall in spatial resolution in the SPECT depends on openings of the collimator 3. The fall in spatial resolution due to the collimator openings depends on a distance between the radiation source and the detector 2 (= the collimator). Thus, the data on the straight line Y=−n/γ in G (γ, n) is subjected to a fall in spatial resolution of the same degree. It is possible to separate the distances d on the frequency space. It is possible to represent the fall in spatial resolution due to the collimator openings with, for example, the Gaussian function. Thus, as indicated by the following expressions, it is possible to correct the fall in spatial resolution due to the collimator openings by multiplying each data on the straight line Y=−n/γ in G(γ,n) by an inverse function (a correction function) provided for each of the distances d. F(γ,n) represents an ideal FDR image without a fall in spatial resolution. H(γ,n) represents a line spread function (an unsharpness function (see FIG. 6A)) represented on the frequency space.G(γ,n)=H(γ,n)×F(γ,n)F(γ,n)=H−1(γ,n)×G(γ,n) Note that, concerning the inverse function, since an infinitely high high-frequency component is restored if a logical form is used, it is difficult to apply the inverse function in a practical use. Thus, a Metz filter (P=3, 5, or 7) or the like obtained by properly cutting a high-frequency component from a filter function of P=∞ is used (see FIG. 6B). The three-dimensional filter processing in the SPEC apparatus in this embodiment is executed using the FDR filter processing. Procedures for the correction processing for a fall in spatial resolution (unsharpness) according to this embodiment are shown in FIG. 3. Imaging (counting) is repeated in plural positions while the detector apparatus 1 intermittently or continuously rotates around the patient P (S11). Collected SPECT data is stored in the image storing unit 9. As shown in FIG. 8A, the sinogram transforming unit 14 expands a one-dimensional projection distribution concerning a detection position X along an axis of a projection angle φ and transforms the one-dimensional projection distribution into a three-dimensional sinogram g(X,φ,Z) shown in FIG. 8B arranged for each slice (S12). The Fourier transformation unit 16 subjects the three-dimensional sinogram g(X,φ,Z) to three-dimensional Fourier transformation concerning three axes X, φ, Z (S13) to obtain a three-dimensional sinogram G(γ,n,w) shown in FIG. 8C represented on a frequency space. γ represents a frequency component in a detection position, n represents a frequency component of an angle (a projection angle) of the detector 3, and w represents a frequency component in a slice direction. The filter processing unit 15 convolutes a correction function for each of the distances d with respect to the three-dimensional sinogram G(γ,n,w) represented on the frequency space (S14). Design of a filter for correcting the fall in spatial resolution due to the collimator openings is described. First, two-dimensionally, in a (γ,n) surface of G(γ,n,w), a filter is designed for each straight line Y=−n/γ on the basis of a logic of FDR. For example, a spatial resolution depending on a distance between a collimator and a radiation source is represented by a mountain-like shape shown in FIG. 6A. A shape of this point spread function can be approximated by the Gaussian distribution, the Poisson distribution, or the like. As the point spread function, a Modulation Transfer Function (MTF) is adopted. The MTF is represented by the following expression.M(f)=eqq=−2π2σ2f2 A position resolution is reflected in a half-value breadth. This distribution is created for each of the distances d between the collimator and the radiation source and set as a response function (S21). An inverse function of the response function created for each of the distances d is created (S22). In other words, a two-dimensional filter is created for each of the distances d. Note that, since the inverse function may diverge at a high frequency, a high-frequency component is cut to a degree not intensifying noise. For example, the inverse function has a shape like a Metz filter shown in FIG. 6B. The Metz filter is represented by the following expression. Note that P is a statistical fluctuation control parameter serving as a parameter determining a filter characteristic. When P increases, the Metz filter has a higher correction effect but is sensitive to noise. When P decreases, the Metz filter has a lower correction effect but has stronger resistance against noise.Metz(f)={1−(1−M(f)2)P}/M(f) Expansion to three dimensions including the slice direction only has to be performed by rotating the correction filter obtained two-dimensionally because data with the same radiation source to detector (collimator) distances including the slice direction are present on a surface including the straight line Y=−n/γ and a straight line of a w axis (S23). Actually, to expand a filtering object from two dimensions to three dimensions, the filter is expanded from one dimension to two dimensions. Specifically, a point spread function shown in FIG. 7A can be expanded to a two-dimensional distribution shown in FIG. 7B by rotating the point spread function conically around an axis passing through an origin (a frequency=0) of an abscissa. As shown in FIG. 9, G(γ,n,w) is corrected by a three-dimensional correction filter serving as an inverse function of a two-dimensional point spread function designed for each surface including the straight line Y−n/γ and the straight line of the w axis. The three-dimensional correction filter is convoluted with respect to respective local areas with respective points of G(γ,n,w) as centers. The three-dimensional correction filter is properly used according to the distances d. Consequently, it is possible to properly correct a fall in spatial resolution according to the distances d. The inverse function of the two-dimensional point spread function, that is, a filter function is created for each of the distances d in advance. Plural filter functions are stored in the storing unit 13 in association with the distances d. Note that, as described above, the filter function has the statistical fluctuation control parameter P. It is effective to properly use plural kinds of filter functions with different parameters P according to a noise level of collected data. When P increases, the filter function has a higher correction effect but is sensitive to noise. When P decreases, the filter function has a lower correction effect but has stronger resistance against noise. Therefore, the plural kinds of filter functions with different parameters P are created in advance. The plural kinds of filter functions with different parameters P are stored in the storing unit 13 together with the distances d in association with the parameters P. The filter processing unit 15 can properly use the filter functions according to the distances d and the parameters P. Typically, an amount of RIs that can be dosed in a patient is substantially determined according to a physique (weight, height, etc.) of the patient, a test region, and a type of a drug marked by the RIs. When a dosage of the RIs is large, noise is small. Conversely, when a dosage of the RIs is small, noise is large. The filter processing unit 15 can select an optimum filter function in accordance with the dosage of the RIs and the test region and the like determining the dosage. A filter function satisfying a condition, with which an influence of noise is not made manifest, and having a highest correction effect, that is, a highest parameter P corresponds to the optimum filter function. The filter processing unit 15 corrects plural two-dimensional projection distributions (a three-dimensional sinogram) initially using the optimum filter function selected in accordance with the dosage of the RIs and the test region and the like determining the dosage. A three-dimensional RI distribution is reconfigured and displayed on the basis of the corrected three-dimensional sinogram. Note that the filter processing unit 15 may correct three-dimensional sinograms initially using all the filter functions with different filter characteristics, that is, different statistical fluctuation control parameters P in this embodiment, respectively. In this case, on the basis of the corrected plural three-dimensional sinograms, RI distributions with different filter characteristics corresponding to the three-dimensional sinograms are reconfigured by the reconfiguration processing unit 18. The plural three-dimensional RI distributions are displayed in a display unit 10 as a list. The inverse Fourier transformation unit 17 returns G(γ,n,w), which has the fall of spatial resolution corrected, to the three-dimensional sinogram g(X,φ,Z) represented on the actual space according to the inverse Fourier transformation (S15). The reconfiguration processing unit 18 reconfigures a SPECT tomogram with a multi-stage surface from the three-dimensional sinogram g(X,φ,Z) that has the fall in spatial resolution corrected and is returned to the representation on the actual space (S16). According to this embodiment, it is possible to separate the radiation source to detector distances d by transforming a two-dimensional projection distribution in multiple directions into a three-dimensional sinogram and transferring the three-dimensional sinogram to a frequency space. Thus, it is possible to effectively correct a fall in spatial resolution with an appropriate correction function according to the distances d. Additional advantages and modifications will readily occur to those skilled in the art. Therefore, the invention in its broader aspects is not limited to the specific details and representative embodiments shown and described herein. Accordingly, various modifications may be made without departing from the spirit or scope of the general inventive concept as defined by the appended claims and their equivalents.
	claims	1. A method comprising:supplying deuterium and 3He to a fusion reactor;conducting fusion reactions including deuterium-3He (D-3He) fusion reactions and deuterium-deuterium (D-D) fusion reactions to produce reaction products comprising 3He and tritium byproducts;pulsing the fusion reactor; andremoving, responsive to the pulsing of the fusion reactor, at least some of the tritium byproducts from the fusion reactor,wherein the at least some of the tritium byproducts from the D-D fusion reactions are removed prior to a deuterium-tritium (D-T) fusion reaction, andwherein the at least some of the tritium byproducts that are removed prior to contributing to a D-T fusion reaction are produced by the fusion reactor during the D-D fusion reactions. 2. The method of claim 1, wherein the tritium is removed from the fusion reactor to decay and create 3He. 3. The method of claim 1, wherein the tritium removed from the reactor is allowed to decay into 3He, which is subsequently supplied to the fusion reactor together with additional deuterium. 4. The method of claim 3, wherein the 3He is used by supplying the 3He to the fusion reactor together with additional deuterium. 5. The method of claim 3, wherein the 3He supplied to the fusion reactor comprises 3He from previous D-D fusion reactions to allow for a self-sustaining D-3He fuel cycle with no external 3He addition. 6. The method of claim 1, further comprising storing the tritium byproducts in a location remote from the fusion reactor. 7. The method of claim 1, wherein conducting the D-D fusion reactions comprises forming at least two plasmoids and accelerating the at least two plasmoids towards one another. 8. The method of claim 1, further comprising providing a lithium blanket for production of additional 3He. 9. The method of claim 1, wherein the tritium byproducts are removed between pulses of the fusion reactor. 10. The method of claim 1, further comprising suppressing the D-D reaction based on a temperature at which the fusion reactions are performed and suppressing the D-T reaction by removing the at least some of the tritium byproducts. 11. The method of claim 1, further comprising performing the fusion reactions at a temperature where a fusion reactivity for D-3He is greater than a fusion reactivity for D-D. 12. The method of claim 1, further comprising removing, responsive to the pulsing of the fusion reactor, at least some of the 3He byproducts. 13. The method of claim 12, further comprising supplying the removed 3He as fuel for subsequent fusion reactions.
	summary
	claims	1. An ion implanter system, the system comprising:an ion beam generator for generating an ion beam;a platen;a first test wafer disposed on the platen toward which the ion beam is directed;a second test wafer disposed on the platen toward which the ion beam is directed along a wall of at least one template, the at least one template positioned adjacent an ion beam path, each template having a template surface that impedes a motion of an ion in the ion beam in response to the ion impacting the template surface, wherein the ion impacts the template surface in the case that a trajectory of the ion varies from an optimum trajectory by at least a pre-determined maximum variance angle and wherein the ion impacts the second test wafer in the case that the trajectory of the ion varies from the optimum trajectory by less than the pre-determined maximum variance angle; andmeans for determining, using the at least one template and the first and second test wafers, whether an ion beam trajectory varies from the optimum trajectory by at least the pre-determined maximum variance angle to monitor an integrity of the ion beam, wherein the means for determining measures a difference between a first number of ions that impact the first test wafer and a second number of ions that impact the second test wafer to determine whether the ion beam trajectory varies from the optimum trajectory by at least the pre-determined maximum variance angle. 2. The system of claim 1, wherein the template surface undergoes a change in response to the impacting by the ion and wherein the means for determining measures the change in the template surface to determine whether the ion beam trajectory varies from the optimum trajectory by at least the pre-determined maximum variance angle. 3. The system of claim 1, wherein the at least one template includes an opening having a template opening width and a template wall having a template wall depth, and wherein an aspect ratio of the template opening width to the template wall depth corresponds to the pre-determined maximum variance angle. 4. The system of claim 3, wherein the at least one template includes a plurality of templates, wherein the aspect ratio of at least one template differs from the aspect ratio of at least one other template. 5. The system of claim 1, wherein the at least one template comprises at least two stacked templates oriented with an edge of the template wall of one template substantially perpendicular to the edge of the template wall of at least one other template. 6. The system of claim 1, wherein at least a portion of the template surface is at least one of a photoresist, an oxide and a nitride hard mask. 7. The system of claim 1, further comprising means for determining an adjustment of the ion beam based on an input from the means for determining. 8. A system for monitoring an integrity of an ion beam, the system comprising:a platen;a first test wafer disposed on the platen toward which an ion beam is directed;a second test wafer disposed on the platen toward which the ion beam is directed along a wall of at least one template, the at least one template positioned adjacent an ion beam path, each template having a template surface that impedes a motion of an ion wherein the ion impacts the template wall in the case that a trajectory of the ion varies from an optimum trajectory by at least a pre-determined maximum variance angle and wherein the ion impacts the second test wafer in the case that the trajectory of the ion varies from the optimum trajectory by less than the pre-determined maximum variance angle; andmeans for determining, using the at least one template and the first and second test wafers, whether an ion beam trajectory varies from the optimum trajectory by at least the pre-determined maximum variance angle to monitor the integrity of the ion beam, wherein the means for determining measures a difference between a first number of ions that impact the first test wafer and a second number of ions that impact the second test wafer to determine whether the ion beam trajectory varies from the optimum trajectory by at least the pre-determined maximum variance angle. 9. The system of claim 8, wherein the template surface undergoes a change in response to the impacting by the ion and wherein means for determining measures the change in the template surface to determine whether the ion beam trajectory varies from the optimum trajectory by at least the pre-determined maximum variance angle. 10. The system of claim 8, wherein the at least one template includes an opening having a template opening width and a template wall having a template wall depth, and wherein an aspect ratio of the template opening width to the template wall depth corresponds to the maximum variance angle. 11. The system of claim 10, wherein the at least one template includes a plurality of templates, wherein the aspect ratio of at least one template differs from the aspect ratio of at least one other template. 12. The system of claim 8, wherein the at least one template comprises at least two stacked templates oriented with an edge of a template wall of one template substantially perpendicular to the edge of the template wall of at least one other template. 13. The system of claim 8, wherein at least a portion of the template surface is at least one of a photoresist, an oxide and a nitride hard mask. 14. The system of claim 8, further comprising means for determining an adjustment of the ion beam based on an input from the means for determining. 15. A method of monitoring angle integrity of an ion beam, the method comprising the steps of:transmitting an ion beam toward a first test wafer disposed on a platen;transmitting the ion beam adjacent to at least one template toward a second test wafer disposed on the platen along a wall of the template, the at least one template having a template surface that impedes a motion of an ion that varies from an optimum trajectory by at least a pre-determined maximum variance angle and wherein the ion impacts the second test wafer in the case that the trajectory of the ion varies from the optimum trajectory by less than the pre-determined maximum variance angle; anddetermining, using the at least one template and the first and second test wafers, whether an ion beam trajectory varies from the optimum trajectory by at least the pre-determined maximum variance angle to monitor angle integrity of the ion beam, wherein the determining step measures a difference between a first number of ions that impact the first test wafer and a second number of ions that impact the second test wafer to determine whether the ion beam trajectory varies from the optimum trajectory by at least the pre-determined maximum variance angle. 16. The method of claim 15, wherein the template surface undergoes a change in response to the impacting by the ion and wherein the determining step measures the change in the template surface to determine whether the ion beam trajectory varies from the optimum trajectory by at least the pre-determined maximum variance angle. 17. The method of claim 15, wherein the at least one template includes an opening having a template opening width and a template wall having a template wall depth, and wherein an aspect ratio of the template opening width to the template wall depth corresponds to the maximum variance angle. 18. The method of claim 17, wherein the at least one template includes a plurality of templates, wherein the aspect ratio of at least one template differs from the aspect ratio of at least one other template. 19. The method of claim 15, wherein the at least one template comprises at least two stacked templates oriented with an edge of the template wall of one template substantially perpendicular to the edge of the template wall of at least one other template. 20. The method of claim 15, wherein at least a portion of the template surface is at least one of a photoresist, an oxide and a nitride hard mask. 21. The method of claim 15, further comprising, determining an adjustment of the angle integrity of the ion beam based on the determining step. 22. A computer program product comprising a computer useable medium having computer readable program code embodied therein for determining the integrity of the angle of an ion beam generated by a ion implanter system and transmitted by the ion implanter system, the program product comprising:program code configured to control transmitting an ion beam toward a first test wafer disposed on a platen;program code configured to control transmitting the ion beam adjacent to at least one template along a wall of the template toward a second test wafer disposed on the platen, the at least one template having a surface that impedes a motion of an ion that varies from an optimum trajectory by at least a pre-determined maximum variance angle and wherein the ion impacts the second test wafer in the case that the trajectory of the ion varies from the optimum trajectory by less than the pre-determined maximum variance angle; andprogram code configured to determine, using the at least one template and the first and second test wafers, whether an ion beam trajectory varies from the optimum trajectory by at least the pre-determined maximum variance angle to monitor angle integrity of the ion beam, wherein the program code measures a difference between a first number of ions that impact the first test wafer and a second number of ions that impact the second test wafer to determine whether the ion beam trajectory varies from the optimum trajectory by at least the pre-determined maximum variance angle. 23. The program product of claim 22, wherein the template surface undergoes a change in response to the impacting by the ion and wherein the program code configured to determine measures the change in the template surface to determine whether an ion beam trajectory varies from the optimum trajectory by at least the pre-determined maximum variance angle. 24. The program product of claim 22, further comprising program code configured to determine an adjustment of the ion beam based on an input from the program code configured to determine.
044328923	abstract	For the safe intermediate and final storage of tritium after reaction with a hydride forming metal, it is embedded in concrete which, of course, does not lead to a complete exclusion from the environment since the tritium is able to permeate through the concrete. This is avoided by pressing to molded bodies at room temperature the tritium containing metal particles with powders of metals which have a low permeability for tritium.
	description	1. Field of the Invention The present invention relates to an assembly for detection of at least one of radiation flux and contamination on an optical component. The present invention also relates to a lithographic apparatus including such an assembly. The present invention also relates to a device manufacturing method. 2. Background of the Invention Commonly assigned, co-pending U.S. application Ser. No. 10/686,813, filed Oct. 17, 2003, discloses an electrode in the vicinity of an optical component, e.g. a mirror, in a lithographic apparatus. The electrode is biased with a positive potential relative to the optical component. Electrons generated from the optical component by a beam of EUV radiation are transported to the electrode. The current caused by the electrons is measured by a measuring device. This current is correlated to the radiation flux of the beam. It is desirable to monitor the radiation flux at different locations and/or the contamination on the optical components in a lithographic projection apparatus. When radiation hits a surface it induces secondary electrons. The flux of the secondary electrons thus generated is related to the radiation flux on the surface and/or contamination of the surface. Presently, the secondary electrons are extracted from the surface by an electrode creating an electron attracting electric field and the current through the optical component is measured. In some instances, this measured current may not always yield the correct value for the secondary electron flux. This is caused by the following phenomenon. An photon that hits the surface induces quite an amount of secondary electrons. These electrons form a space charge layer in front of the surface of the optical component for a substantial period of time, since a large flux of secondary electrons is created and since it takes the secondary electrons some time to travel to the electrode. The space charge cloud is located between the surface and the electrode. Newly created secondary electrons are obstructed by this space charge cloud which in a way “screens” the electrode such that less secondary electrons than are generated arrive at the electrode. This may result in an inadequate assessment of the radiation flux. It is an aspect of the present invention to provide a device that establishes the correct value of the radiation dose more accurately. The present invention according to a first embodiment includes a detector including at least one wire. The detector can be placed in an radiation beam without substantially blocking the beam. The wire generates a minimal amount of shadow in the radiation beam. Due to its limited exposure to radiation, heating of the detector is prevented. In addition, the detector makes on-line in-situ measurements possible. In another embodiment, the present invention includes a meter connected to a memory that stores the detector signal as function of time. This makes it possible to discriminate between short time and long time changes as a function of time. Short time changes will likely be caused by changes in the power of the radiation source supplying the radiation flux. On the other hand, long term changes will most probably be caused by contamination of the optical component. In yet another embodiment of the present invention, the detector signal is related to a current in the wire and the meter includes a current meter. The current is a reliable measure for the amount of electrons generated, and hence, of the radiation flux. In addition, the electron flux is proportional to the radiation flux up to a much higher maximum radiation flux. In a further embodiment of the present invention, the assembly is located in the vicinity of the component. In this way, an accurate radiation flux detector is obtained that may also be used to detect contamination of an optical component. In a still further embodiment of the present invention, the wire is negatively biased. By negatively biasing the wire relative to the surroundings of the wire, the generated secondary electrons are swiftly removed in front of the detector, because the small geometry of the wire results in a high electric field at the surface of the wire. A separate electrode may function as counter electrode to the wire, or the optical component itself may serve as a counter electrode. In a still further embodiment of the present invention, the wire includes at least one dielectric material and the wire forms a portion of a closed loop connected to a voltage source. Such an installation shows a quick response to incident radiation flux. It is feasible that by choosing different dielectric materials it is possible to distinguish between different contaminants. In still a further embodiment of the present invention, the wire is at least partly covered with a conducting layer. As an example, for the conducting layer a thin (0.1-100 nm) silicon (Si) layer may be chosen. The silicon oxidizes to silicon dioxide (SiO2). The conductivity of silicon is different from the conductivity of silicon dioxide. As the conducting layer is present on a wire made of a dielectric (i.e. insulating) material, this embodiment has the advantage that the conductivity of the wire is significantly influenced by on changes in the conductivity of the conducting layer. The changes in conductivity relate to the amount of oxidation of the conducting layer. In still a further embodiment of the present invention, the wire is placed in the vicinity of a radiation source to measure an amount of debris ejected by the radiation source and the wire is connected to a voltage source. This assembly is particularly suited to detect radiation source induced emission of contaminating particles. In still a further embodiment of the present invention, the wire is located on an optical component. This has the advantage of an accurate measurement with respect to position on an optical component. In addition, the optical component will act as a support. This means that the wire can be much thinner than when the wire is not located on an optical component. Hence, the obstruction of the radiation is less severe. In a still further embodiment of the present invention, the optical component comprises a multilayer mirror. This allows damage to a layer of a multilayer mirror to be detected early. In still a further embodiment of the present invention, the wire includes at least one fluorescent portion. In this way, the high energy (e.g., UV or EUV) radiation which may damage a meter may be converted to alternative radiation (or fluorescent). This alternative radiation also may be easier to detect. In addition, since the alternative radiation will be emitted in a random direction, this alternative radiation may be detected with a detector outside the radiation beam. In still a further embodiment of the present invention, the detector includes a plurality of wires electrically isolated with respect to each other and forming a mesh and, a plurality of meters to measure an individual detector signals, when the assembly is in use, generated by each of the plurality of wires. This provides a radiation flux measurement and contamination detection that are spatially resolved. In still a further embodiment of the present invention, the voltage source, when the assembly is in use, supplies a voltage in synchronism with the radiation flux. This allows application of a voltage only when required, i.e. during the presence of radiation. Voltages do have an effect on charged particles in the lithographic apparatus. This effect is reduced in this embodiment. In still a further embodiment of the present invention, at least one wire is at least partly enclosed by a shielding device. This reduces the effect of external influences on the wire. According to another aspect of the present invention, a lithographic apparatus includes an illumination system configured to provide a projection beam of radiation; a support configured to support a patterning structure, the patterning structure configured to impart the projection beam with a pattern in its cross-section; a substrate table configured to hold a substrate; and a projection system configured to project the patterned beam onto a target portion of the substrate, wherein the lithographic projection apparatus includes an assembly as described above. A lithographic apparatus is a machine that applies a desired pattern onto a target portion of a substrate. Lithographic apparatus can be used, for example, in the manufacture of integrated circuits (ICs). In that circumstance, a patterning structure, such as a mask, may be used to generate a circuit pattern corresponding to an individual layer of the IC, and this pattern can be imaged onto a target portion (e.g. comprising part of, one or several dies) on a substrate (e.g. a silicon wafer) that has a layer of radiation-sensitive material (resist). In general, a single substrate will contain a network of adjacent target portions that are successively exposed. Known lithographic apparatus include so-called steppers, in which each target portion is irradiated by exposing an entire pattern onto the target portion in one go, and so-called scanners, in which each target portion is irradiated by scanning the pattern through the projection beam in a given direction (the “scanning”-direction) while synchronously scanning the substrate parallel or anti-parallel to this direction. According to another aspect of the present invention, a device manufacturing method includes projecting a patterned beam of radiation onto a target portion of the substrate, and detecting at least one of radiation flux and contamination of an optical component by generating a signal correlated to at least one of the radiation flux and the contamination and measuring the signal. Although specific reference may be made in this text to the use of lithographic apparatus in the manufacture of ICs, it should be understood that the lithographic apparatus described herein may have other applications, such as the manufacture of integrated optical systems, guidance and detection patterns for magnetic domain memories, liquid-crystal displays (LCDs), thin-film magnetic heads, etc. One of ordinary skill will appreciate that, in the context of such alternative applications, any use of the terms “wafer” or “die” herein may be considered as synonymous with the more general terms “substrate” or “target portion”, respectively. The substrate referred to herein may be processed, before or after exposure, in for example a track (a tool that typically applies a layer of resist to a substrate and develops the exposed resist) or a metrology or inspection tool. Where applicable, the disclosure herein may be applied to such and other substrate processing tools. Further, the substrate may be processed more than once, for example in order to create a multi-layer IC, so that the term substrate used herein may also refer to a substrate that already contains multiple processed layers. The terms “radiation” and “beam” used herein encompass all types of electromagnetic radiation, including ultraviolet (UV) radiation (e.g. having a wavelength of 365, 248, 193, 157 or 126 nm) and extreme ultra-violet (EUV) radiation (e.g. having a wavelength in the range of 5-20 nm), as well as particle beams, such as ion beams or electron beams. The term “patterning structure” used herein should be broadly interpreted as referring to structure that can be used to impart a projection beam with a pattern in its cross-section such as to create a pattern in a target portion of the substrate. It should be noted that the pattern imparted to the projection beam may not exactly correspond to the desired pattern in the target portion of the substrate. Generally, the pattern imparted to the projection beam will correspond to a particular functional layer in a device being created in the target portion, such as an integrated circuit. Patterning structure may be transmissive or reflective. Examples of patterning structures include masks, programmable mirror arrays, and programmable LCD panels. Masks are well known in lithography, and include mask types such as binary, alternating phase-shift, and attenuated phase-shift, as well as various hybrid mask types. An example of a programmable mirror array employs a matrix arrangement of small mirrors, each of which can be individually tilted so as to reflect an incoming radiation beam in different directions; in this manner, the reflected beam is patterned. The support supports, i.e. bears the weight of, the patterning structure. It holds the patterning structure in a way depending on the orientation of the patterning structure, the design of the lithographic apparatus, and other conditions, such as for example whether or not the patterning structure is held in a vacuum environment. The support can use mechanical clamping, vacuum, or other clamping techniques, for example electrostatic clamping under vacuum conditions. The support structure may be a frame or a table, for example, which may be fixed or movable as required and which may ensure that the patterning structure is at a desired position, for example with respect to the projection system. Any use of the terms “reticle” or “mask” herein may be considered synonymous with the more general term “patterning structure”. The term “projection system” used herein should be broadly interpreted as encompassing various types of projection system, including refractive optical systems, reflective optical systems, and catadioptric optical systems, as appropriate for example for the exposure radiation being used, or for other factors such as the use of an immersion fluid or the use of a vacuum. Any use of the term “lens” herein may be considered as synonymous with the more general term “projection system”. The illumination system may also encompass various types of optical components, including refractive, reflective, and catadioptric optical components for directing, shaping, or controlling the projection beam of radiation, and such components may also be referred to below, collectively or singularly, as a “lens.” The lithographic apparatus may be of a type having two (dual stage) or more substrate tables (and/or two or more mask tables). In such “multiple stage” machines the additional tables may be used in parallel, or preparatory steps may be carried out on one or more tables while one or more other tables are being used for exposure. The lithographic apparatus may also be of a type wherein the substrate is immersed in a liquid having a relatively high refractive index, e.g. water, so as to fill a space between the final element of the projection system and the substrate. Immersion liquids may also be applied to other spaces in the lithographic apparatus, for example, between the mask and the first element of the projection system. Immersion techniques are well known in the art for increasing the numerical aperture of projection systems. In the drawings, corresponding reference symbols indicate corresponding parts. FIG. 1 schematically depicts a lithographic apparatus according to the present invention. The apparatus includes an illumination system (illuminator) IL configured to provide a projection beam PB of radiation (e.g. UV or EUV radiation). A first support (e.g. a mask table) MT is configured to support a patterning structure (e.g. a mask) MA and is connected to a first positioning device PM that accurately positions the patterning structure with respect to a projection system (lens) PL. A substrate table (e.g. a wafer table) WT is configured to hold a substrate (e.g. a resist-coated wafer) W and is connected to a second positioning device PW that accurately positions the substrate with respect to the projection system PL. The projection system (e.g. a reflective projection lens) PL images a pattern imparted to the projection beam PB by the patterning structure MA onto a target portion C (e.g. including one or more dies) of the substrate W. As here depicted, the apparatus is of a reflective type (e.g. employing a reflective mask or a programmable mirror array of a type as referred to above). Alternatively, the apparatus may be of a transmissive type (e.g. employing a transmissive mask). The illuminator IL receives a beam of radiation from a radiation source SO. The source and the lithographic apparatus may be separate entities, for example when the source is a plasma discharge source. In such cases, the source is not considered to form part of the lithographic apparatus and the radiation beam is generally passed from the source SO to the illuminator IL with the aid of a radiation collector comprising for example suitable collecting mirrors and/or a spectral purity filter. In other cases the source may be integral part of the apparatus, for example when the source is a mercury lamp. The source SO and the illuminator IL, may be referred to as a radiation system. The illuminator IL may comprise an adjusting device configured to adjust the angular intensity distribution of the beam. Generally, at least the outer and/or inner radial extent (commonly referred to as σ-outer and σ-inner, respectively) of the intensity distribution in a pupil plane of the illuminator can be adjusted. The illuminator provides a conditioned beam of radiation, referred to as the projection beam PB, having a desired uniformity and intensity distribution in its cross-section. The projection beam PB is incident on the mask MA, which is held on the mask table MT. Being reflected by the mask MA, the projection beam PB passes through the lens PL, which focuses the beam onto a target portion C of the substrate W. With the aid of the second positioning device PW and a position sensor IF2 (e.g. an interferometric device), the substrate table WT can be moved accurately, e.g. so as to position different target portions C in the path of the beam PB. Similarly, the first positioning device PM and a position sensor IF1 can be used to accurately position the mask MA with respect to the path of the beam PB, e.g. after mechanical retrieval from a mask library, or during a scan. In general, movement of the object tables MT and WT will be realized with the aid of a long-stroke module (coarse positioning) and a short-stroke module (fine positioning), which form part of the positioning device PM and PW. However, in the case of a stepper (as opposed to a scanner) the mask table MT may be connected to a short stroke actuator only, or may be fixed. Mask MA and substrate W may be aligned using mask alignment marks M1, M2 and substrate alignment marks P1, P2. The depicted apparatus can be used in the following preferred modes: 1. In step mode, the mask table MT and the substrate table WT are kept essentially stationary, while an entire pattern imparted to the projection beam is projected onto a target portion C in one go (i.e. a single static exposure). The substrate table WT is then shifted in the X and/or Y direction so that a different target portion C can be exposed. In step mode, the maximum size of the exposure field limits the size of the target portion C imaged in a single static exposure.2. In scan mode, the mask table MT and the substrate table WT are scanned synchronously while a pattern imparted to the projection beam is projected onto a target portion C (i.e. a single dynamic exposure). The velocity and direction of the substrate table WT relative to the mask table MT is determined by the (de)magnification and image reversal characteristics of the projection system PL. In scan mode, the maximum size of the exposure field limits the width (in the non-scanning direction) of the target portion in a single dynamic exposure, whereas the length of the scanning motion determines the height (in the scanning direction) of the target portion.3. In another mode, the mask table MT is kept essentially stationary holding a programmable patterning structure, and the substrate table WT is moved or scanned while a pattern imparted to the projection beam is projected onto a target portion C. In this mode, generally a pulsed radiation source is employed and the programmable patterning structure is updated as required after each movement of the substrate table WT or in between successive radiation pulses during a scan. This mode of operation can be readily applied to maskless lithography that utilizes a programmable patterning structure, such as a programmable mirror array of a type as referred to above. Combinations and/or variations on the above described modes of use or entirely different modes of use may also be employed. Referring to FIG. 2, a wire 25 is in a projection beam of radiation 21 (this projection beam is referred to in FIG. 1 with “PB” and is preferably, but not necessarily, EUV radiation). The wire 25 is connected to a current meter 27, which meter is grounded, i.e. connected to earth via a connection 29. A microprocessor unit 26 and a memory 28 for storing the measurements of the meter 27 as a function of time may also be connected to the meter 27. The projection beam of radiation 21 is incident on an optical component 23 such as a mirror. Electrons 31 are ejected from the wire 25. The projection beam of radiation 21 that hits the wire 25 causes electrons 31 to be ejected from the wire 25. To maintain the neutral state of the wire 25, the ejected electrons 31 are compensated by electrons from ground. These electrons will pass through the connection 29 and through the current meter 27. The current detected by the meter 27 is correlated to the radiation flux on the wire 25 by the microprocessor unit 26. The microprocessor unit 26 converts the current through wire 25 to a radiation flux. As the wire 25 is located close to optical component 23, the radiation flux that is detected by the assembly of the wire 25 and the meter 27 will approximate and be representative for the radiation flux on the optical component 23. The radiation flux detected by the wire 25 can be used to obtain an indication of the contamination of the optical component 23 in the following way. A decrease in the detected radiation flux in time can be observed by use of the memory 28. If, however, it can be established that the radiation flux of the projection beam of radiation 21 is steady, this would indicate contamination of the wire 25. Since the optical component 23 is located in the vicinity of the wire 25, one may then assume that there is contamination on the optical component 23 too. The thin wire (or a mesh/grid of thin wires, as explained in connection with FIG. 9b) only marginally disrupts the projection beam of radiation, i.e. the shadow caused by such a relatively thin object in the projection beam is marginal. The memory 28 enables time dependent measurements. Referring to FIG. 3, several wires 25 form a mesh 33. The wires 25 are electrically isolated with respect to each other. A counter electrode (shown in FIG. 3 as, for example, a corresponding mesh) 35 is present, outside the projection beam of radiation 21. The counter electrode 35, however, can have any shape. A voltage source 37 is connected in between the mesh 33 and the counter electrode 35. The voltage source 37 can in principle be connected to all the wires 25 of the mesh 33. In order to measure current through single wires, the wires should be connected to separate current meters. This means that several current meters will be present. In FIG. 3, only one wire is shown connected to the current meter 27. However, it is to be understood, although they are not shown in FIG. 3, that corresponding voltage sources are connected to the other wires 25 forming the mesh 33. The voltage source 37 induces a negative potential on the wires 25 of the mesh 33 and a positive potential on the counter electrode 35. Negative potential and positive potential are to be understood as negative and positive in relation to each other. It is sufficient that the counter electrode 35 is on a higher potential than (the wires 25 of) the mesh 33. This means that the counter electrode 35 can have a negative, zero or positive voltage. Electrons 31 generated by the projection beam of radiation 21 are transported from the wires 25 to the counter electrode 35 on a positive potential. The mesh 33 has a smaller surface and a correspondingly high electric field at the surface (when a negative voltage with respect to the surroundings is applied). Furthermore, by isolating the wires 25 of the mesh 33 from each other, a spatially resolved measurement of the dose and contamination can be performed. When operating with a radiation source that provides a pulsed projection beam of radiation, it is desirable to provide the difference in voltage by the voltage source 37 on the mesh 33 and the counter electrode 35 in phase with the pulsed projection beam, i.e. only apply the voltage when beam 21 is present and electrons 31 are generated and no voltage when there is no beam 21. To control this synchronization process a controller 24 is connected to the current meter 27 and the voltage source 37. The controller 24 also receives, via a link 30, information about the status (on/off) of the radiation source. Referring to FIG. 4, the function of the counter electrode 35 is taken over by the optical component 23. Although this may lead to a certain amount of contamination of the optical component 23 by electrons, this amount will generally be small as the amount of electrons 31 generated from the mesh 33 will be small due to the effect of the small surface areas of the mesh 33. Referring to FIG. 5, the wire 25 is, at least partly, made of a dielectric material 26 in such a way that the edges of the wire are not electrically connected. The dielectric may be shaped differently than as a wire. At both edges of the wires an electrode 38, 39 is connected to voltage source 37. A non-dielectric material 47 (for instance in the form of a layer) is present on the dielectric material 26 of the wire 25. As the projection beam of radiation 21 impinges on the non-dielectric material 47, the non-dielectric material 47 can be oxidised. An example of the non-dielectric material 47 is silicon (Si). Upon illumination with radiation, silicon oxidises to SiO2. As the conductivity of the wire 25 is determined by the conductivity of the non-dielectric material 47, the conductivity of the wire changes due to this oxidation process. Therefore, the current through the wire 25 will change as a function of the amount of oxidation of the non-dielectric material 47. Hence, the current as measured by current meter 27 can be related to the oxidation of the non-dielectric material 47. This allows for the determination of oxidation by the projection beam. Referring to FIG. 6, the wire previously described in FIG. 5 is shown, but without the presence of the non-dielectric material 47. Both terminals of the wire in FIG. 6 and in FIGS. 7 and 8 are connected to an assembly similar to the one described above. The wire 25 in FIG. 6 is at least partly made of dielectric material 26 in such a way that the edges of the wire 25 are not electrically connected. As the projection beam of radiation 21 impinges on dielectric material 26, the surface of the dielectric material gets conductive. Therefore, a current will be detected by current meter 27. The current produced is a function of the type of radiation, type of dielectric material, and the contamination of the surface of the dielectric material 26. By choosing several wires 25 with different dielectric materials 26 it is possible to discriminate between different contaminants. Examples of dielectric materials are SiO2 and Al2O3. Other materials with different affinity to different contaminants may be used. This allows for an in-situ contamination determination which distinguishes between contaminants. Referring to FIG. 7, the insulating wire 25 is covered with a conducting layer 40. During operation, the layer 40 will be etched away by ions 39 in the projection beam of radiation 21 of the lithographic apparatus. This again will change the electrical conductivity of the wire 25 as measured by the assembly, including the current meter 27 and the voltage source 37. The assembly measures the influence of accelerated ions in the lithographic apparatus. When the insulating wire 25 with the conducting layer 40 is placed in the vicinity of the radiation source SO, the assembly measures source induced etching (sputtering). Referring to FIG. 8, the wire 25 may also be placed near the radiation source SO (in or outside of the projection beam of radiation 21). In this way, the wire will be covered with ions 39, including sputtered electrode material which also will change its electrical conductivity. Referring to FIG. 9a, instead of a wire 25 separate from an optical component 23, the optical surface of the optical component 23, such as a mirror, is used for contamination detection. Parts 25 of the mirror 23 function in a way comparable to the wire 25 from the above embodiments (that is why they are referred to with the same reference numeral 25). These parts 25 are connected to a voltage source 37 and the current flowing in these parts 25 is measured by the current meter 27. SiO2 or SiC may be used for the insulating layer. FIG. 9b is a front view of the optical component 23 of FIG. 9a. The parts 25 form a network or pattern on the optical component 23. Various different configurations networks, or patterns may be used. Unlike free standing wires, the parts 25 do not have to be as big, since the optical surface acts as support structure for the pattern. Referring to FIG. 10, the optical component 23 is a multilayer mirror comprising several layers 41. The top layer performs a function comparable to the function of the wire 25 described previously, and therefore, is referenced with this same reference numeral. A top layer 25 made of insulating material on the layers 41 may be used for the detection of carbon growth (molecular contamination). With or without the presence of the insulating top layer 25 on the layer 41, the device may be used to detect sputtering of the optical component (multilayer mirror). If the layer 25 is damaged by ions 39, the electrical conductivity will change. Sputtering will also influence conductivity independent of the presence or absence of the layer 25. Referring to FIG. 11, an optical component 23 with a contaminating layer 43 is shown. Contact with the layer 43 is made via removable contacts 44. These contacts 44 are not present during operation of the lithographic apparatus. This embodiment is used to determine ex-situ, (off-line) the amount of contamination on the optical component 23. The arrangement of FIG. 12 is intended to measure radiation flux closely to the reticle level. In FIG. 12, the projection beam of radiation 21 is shown in a cross section. The detection wire 25 is positioned in the projection beam of radiation 21. A screening device 45, such as a shield made of metal, are applied around the wire 25. The inside of the shield is formed of metallic material for conduction reasons. The outside may be of other materials. The structure of the screening device 45 is such that the influence of external fields (e.g. electrical) is minimized. The wire 25 acts an anode while the screening device 45 acts as a cathode. Additionally and alternatively, a metal shield may be positioned around the complete assembly of anode and cathode. The current generated by the radiation flux is measured by a suitable measurement device. By proper geometrical configuration, it is possible to measure the integral of the radiation flux over the complete cross section of the beam. One function of the shielding device is to minimize electrical fields. In case a magnetic field disturbs the measurement, a magnetic shield may be added. Referring to FIG. 13, a part of the surface of the wire 25 includes a layer of fluorescent material 112. The material is sensitive to radiation, for example, EUV radiation. A part 108 of the projection beam 21 will be blocked by the wire 25. A sensor 102 is positioned near the wire, but outside the projection beam 21. The blocked part 108 of projection beam 21 that impinges upon layer 112 of fluorescent material generates light by fluorescence. A portion 104 of that light will impinge upon sensor 102. The sensor 102 measures portion 104 and produces an output signal that is indicative of the intensity of the radiation flux at the position of the wire 25. The sensor 102 may be a photodiode sensitive for visible light. By using fluorescent material, the projection beam profile and intensity can be measured using a relatively simple photodiode. The layer 112 of the wire 25 may, for example comprise compounds like CaS:Ce, YAG:Ce or ZnS:Ag,Al. This arrangement can also be used in a blocking system, such as ReMa blades. Such a blocking system is used in a lithographic apparatus to block off certain parts of the projection beam 21. As the radiation thus blocked would otherwise be lost for further projection purposes, this arrangement may be placed on the blocking system. The blocking system then performs similar to the wire 25. Although the previous descriptions have been explained in the context a wire, the description are equally applicable for a collections of wires forming a mesh or grid. While specific embodiments of the invention have been described above, it will be appreciated that the invention may be practiced otherwise than as described. The description is not intended to limit the invention.
059303204	claims	1. A support assembly for a reactor pressure vessel of a nuclear reactor, the nuclear reactor including a reactor pressure vessel support structure, said support assembly comprising: a support block coupled to the reactor pressure vessel; and a guide block coupled to the reactor pressure vessel support structure, said support block movably coupled to said guide block. a reactor pressure vessel; a reactor pressure vessel support structure; and at least one support assembly coupled between said reactor pressure vessel and said reactor pressure vessel support structure, said support assembly comprising a support block coupled to said reactor pressure vessel, and a guide block coupled to said reactor pressure vessel support structure, said support block movably coupled to said guide block. positioning the support assembly between a portion of the reactor pedestal and a portion of the reactor pressure vessel; and coupling the support block to the reactor pressure vessel. 2. A support assembly in accordance with claim 1 wherein said support block is rigidly coupled to the reactor pressure vessel. 3. A support assembly in accordance with claim 1 wherein said guide block is rigidly coupled to the reactor pressure vessel support structure. 4. A support assembly in accordance with claim 1 wherein said guide block includes a channel sized to receive a portion of said support block. 5. A support assembly in accordance with claim 4 wherein said guide block channel is substantially "T" shaped. 6. A support assembly in accordance with claim 1 further comprising at least one seismic isolator coupled between said guide block and said support block. 7. A support assembly in accordance with claim 1 wherein the reactor pressure vessel includes a flange extending therefrom, and wherein said support block is coupled to said flange. 8. A support assembly in accordance with claim 7 wherein said flange has a substantially segmented ring shape including a plurality of flange portions, and wherein said support block is coupled to one of said flange portions. 9. A nuclear reactor comprising: 10. A nuclear reactor in accordance with claim 9 wherein said support block is rigidly coupled to said reactor pressure vessel. 11. A nuclear reactor in accordance with claim 9 wherein said guide block is rigidly coupled to said reactor pressure vessel support structure. 12. A nuclear reactor in accordance with claim 9 wherein said guide block includes a channel sized to receive a portion of said support block. 13. A nuclear reactor in accordance with claim 12 wherein said channel is substantially "T" shaped. 14. A nuclear reactor in accordance with claim 9 wherein said support assembly further comprises at least one seismic isolator coupled between said guide block and said support block. 15. A nuclear reactor in accordance with claim 9 wherein said reactor pressure vessel comprises a flange extending therefrom, and wherein said support assembly is coupled to said flange. 16. A nuclear reactor in accordance with claim 9 comprising at least two support assemblies, each said support assembly coupled between said reactor pressure vessel and said reactor pressure vessel support structure. 17. A nuclear reactor in accordance with claim 16 wherein said support assemblies are spaced substantially equidistantly about a circumference of said reactor pressure vessel. 18. A method of reducing stress on a nuclear reactor pressure vessel utilizing a support assembly, the nuclear reactor pressure vessel positioned adjacent a reactor pedestal, the support assembly including a support block movably coupled to a guide block, said method comprising the steps of: 19. A method in accordance with claim 18 further comprising the step of coupling the guide block to the reactor pedestal.
	summary
	abstract	An automated machine for loading nuclear reactor fuel, BP, or APSR pellets from small trays having rows of pellets into reactor rods of long length wherein the pellets are individually counted and loaded into rows of a loading tray according to the desired number of pellets per row and total number of pellets per tray with the loading tray controls being moved to a que tray where they may be weighted and there from to an unloading tray for sequential row loading into the reactor rod.
	claims	1. A radiation shield comprising:a first layer comprising a neutron moderating material;a second layer adjacent the first layer, wherein the second layer comprises a particulate neutron absorbing material dispersed in a polymeric binder, wherein the particulate neutron absorbing material comprises at least one neutron absorbing material selected from the group consisting of gadolinium, a gadolinium compound, boron, and a boron compound;a third layer adjacent the second layer, wherein the third layer comprises a photonic radiation attenuating material; andwherein at least one of the first layer and the second layer are removable from the radiation shield. 2. The radiation shield of claim 1, wherein the second layer is intermediate the first layer and the third layer. 3. The radiation shield of claim 1, wherein the neutron moderating material of the first layer comprises a hydrogen-rich polymer. 4. The radiation shield of claim 1, wherein the neutron moderating material of the first layer comprises polyethylene. 5. The radiation shield of claim 1, wherein the third layer is removable from the radiation shield. 6. The radiation shield of claim 1, wherein the first layer is bonded to the second layer, and wherein the first and second layers are removable from the radiation shield as a single unit. 7. The radiation shield of claim 1, wherein the polymeric binder includes at least one material selected from the group consisting of a polyolefin, a polyamide, a polyester, a silicone, a thermoplastic elastomer, and an epoxy. 8. The radiation shield of claim 1, wherein the second layer comprises a layer of neutron absorbing metal or alloy. 9. The radiation shield of claim 1, wherein the second layer comprises a layer of at least one of a neutron absorbing gadolinium alloy and a neutron absorbing boron alloy. 10. The radiation shield of claim 9, wherein the alloy further comprises at least one of copper and aluminum. 11. The radiation shield of claim 1, wherein the second layer comprises one of a metal or alloy layer that is at least one of rolled and cast. 12. The radiation shield of claim 1, wherein the third layer comprises a particulate photonic radiation attenuating material dispersed in a second polymeric binder. 13. The radiation shield of claim 12, wherein the particulate photonic radiation attenuating material comprises tungsten. 14. The radiation shield of claim 12, wherein the second polymeric binder includes at least one material selected from the group consisting of a polyolefin, a polyamide, a polyester, a silicone, a thermoplastic elastomer, and an epoxy. 15. The radiation shield of claim 1, wherein the third layer comprises a tungsten heavy alloy. 16. A device for attenuating radiation comprising at least a first radiation shield panel, the first radiation shield panel comprising:a first layer comprising a neutron moderating material;a second layer adjacent the first layer, wherein the second layer comprises a particulate neutron absorbing material dispersed in a polymeric binder, wherein the particulate neutron absorbing material comprises at least one neutron absorbing material selected from the group consisting of gadolinium, a gadolinium compound, boron and a boron compound; anda third layer adjacent the second layer, wherein the third layer comprises a photonic radiation attenuating material;and wherein at least one of the first layer and the second layer are removable from the first radiation shield panel. 17. The device of claim 16, further comprising a second radiation shield panel, wherein the first radiation shield panel comprises a first edge and the second radiation shield panel comprises a second edge, and wherein the first edge and the second edge include interlocking features forming an interface between the first radiation shield panel and the second radiation shield panel. 18. An apparatus comprising:a radiation-emitting source; anda radiation shield adjacent the radiation-emitting source, the radiation shield comprising:a first layer comprising a neutron moderating material;a second layer adjacent the first layer, wherein the second layer comprises a particulate neutron absorbing material dispersed in a polymeric binder, wherein the particulate neutron absorbing material comprises at least one neutron absorbing material selected from the group consisting of gadolinium, a gadolinium compound, boron, and a boron compound; anda third layer adjacent the second layer, wherein the third layer comprises a photonic radiation attenuating material;and wherein at least one of the first layer and the second layer are removable from the radiation shield panel. 19. A method of shielding an object from a radiation source, the method comprising:placing a radiation shield intermediate the object and the radiation source, wherein the radiation shield comprises:a first layer comprising a particulate neutron absorbing material dispersed in a polymeric binder, wherein the particulate neutron absorbing material comprises at least one neutron absorbing material selected from the group consisting of gadolinium, a gadolinium compound, born, and a boron compound, anda second layer comprising a photonic radiation attenuating material;monitoring the neutron transmissivity of the radiation shield; andreplacing at least a portion of the first layer when the neutron transmissivity of the radiation shield exceeds a predetermined value. 20. The method of claim 19, wherein the second layer further comprises a neutron moderating material. 21. A radiation shield comprising:a first layer comprising a particulate neutron absorbing material dispersed in a polymeric binder, wherein the particulate neutron absorbing material comprises at least one neutron absorbing material selected from the group consisting of gadolinium, a gadolinium compound, boron, and a boron compound;a second layer adjacent the first layer, wherein the second layer Comprises a photonic radiation attenuating material; andwherein the first layer is removable from the radiation shield. 22. The radiation shield of claim 21, wherein the polymeric binder comprises a hydrogen rich polymer. 23. The radiation shield of claim 22 wherein the hydrogen rich polymer includes polyethylene. 24. The radiation shield of claim 21, wherein the polymeric binder includes at least one material selected from the group consisting of a polyolefin, a polyamide, a polyester, a silicone, a thermoplastic elastomer, and an epoxy. 25. The radiation shield of claim 21, wherein the third layer comprises a tungsten heavy alloy. 26. The radiation shield of claim 21, wherein the third layer comprises a particulate photonic radiation attenuating material dispersed in a second polymeric binder. 27. The radiation shield of claim 26, wherein the particulate photonic radiation attenuating material comprises tungsten.
050680811	abstract	In a method of assembling a nuclear fuel assembly, a deflecting jig is inserted into grid cells in each of a plurality of grids. The diameter of the deflecting jig is enlarged to urge a spring of at least one pair of dimples and spring associated with the grid cell to deflect the spring away from the dimple. A plurality of elongated key members are inserted into the grid cells through a plurality of openings defined at intersections between the straps forming walls of the grid cells. Each key member is rotated about its axis to cause hooks of the key member to project from a wall surface of the strap in a direction opposite to the projecting direction of the springs. The key member is then moved in a direction to engage the hooks with the wall surface of the strap. Urging of the spring by the deflecting jig is released to allow the same to be withdrawn from the grid cells and, subsequently, the fuel rods are inserted into the respective grid cells. The key number is then moved in a direction to bring the springs into pressure contact with the fuel rods. The key members are then withdrawn from the grid cells.
	summary
	claims	1. A support apparatus that is usable with a number of springs of a spring apparatus of a nuclear installation, the spring apparatus having a plurality of elongated springs that are stacked together one upon the other and that are affixed at a location thereon to a nozzle of a fuel assembly of the nuclear installation, the number of springs at another location thereon that is spaced from the location being engaged with an upper core plate of the nuclear installation, the number of springs engaged between the nozzle and the upper core plate being deformed between a compressed state and another compressed state when the nuclear installation is operated between a cold condition and a hot condition, the compressed state and the another compressed state being different than one another, the support apparatus comprising:a support that is structured to be stacked together with the number of springs and to be affixed to the nozzle; andan abutment apparatus that comprises a bumper that is affixed to the support and that is spaced a first distance from an end of the support, the bumper protruding a second distance away from a surface of the support and being structured to engage the number of springs at a position on the number of springs disposed between the location and the another location during at least a portion of the deformation between the compressed state and the another compressed state. 2. The support apparatus of claim 1 wherein the abutment apparatus further comprises another bumper that is affixed to the support and that is spaced another first distance from the end, the another bumper protruding another second distance away from the surface of the support and being structured to engage the number of springs at another position on the number of springs disposed between the location and the another location during another portion of the deformation between the compressed state and the another compressed state. 3. The support apparatus of claim 2 wherein at least one of:the another first distance is greater than the first distance; andthe another second distance is greater than the second distance. 4. The support apparatus of claim 2 wherein the another first distance is greater than the first distance, and the another second distance is greater than the second distance. 5. The support apparatus of claim 2 wherein:the plurality of springs are in the compressed state at the cold condition and are in the another compressed state at the hot condition; andthe bumper and the another bumper are structured to simultaneously engage the number of springs during the deformation between the compressed state and an intermediate compressed state that is between the compressed state and the another compressed state, the compressed state and the another compressed state each being different than the intermediate compressed state. 6. The support apparatus of claim 5 wherein the bumper is structured to engage the number of springs and the another bumper is structured to be disengaged from the number of springs during the deformation between the intermediate compressed state and the another compressed state. 7. A spring apparatus that is structured for use in a nuclear installation, the spring apparatus comprising:a number of elongated springs;a support apparatus comprising a support and an abutment apparatus that is situated on the support;the number of springs and the support being stacked together one upon the other and being structured to be affixed at a location thereon to a nozzle of a fuel assembly of the nuclear installation, the number of springs at another location thereon that is spaced from the location being structured to be engaged with an upper core plate of the nuclear installation, the number of springs engaged between the nozzle and the upper core plate being deformed between a compressed state and another compressed state when the nuclear installation is operated between a cold condition and a hot condition, the compressed state and the another compressed state being different than one another; andthe abutment apparatus comprising a bumper that is affixed to the support and that is spaced a first distance from an end of the support, the bumper protruding a second distance away from a surface of the support and being structured to engage the number of springs at a position on the number of springs disposed between the location and the another location during at least a portion of the deformation between the compressed state and the another compressed state. 8. A spring apparatus that is structured for use in a nuclear installation, the spring apparatus comprising:a plurality of elongated springs that are stacked together one upon the other and that are affixed at a location thereon to a nozzle of a fuel assembly of the nuclear installation;when the nuclear installation is in a cold condition, the plurality of springs being in a compressed state and each being compressively engaged at another location thereon that is spaced from the location with an upper core plate of the nuclear installation; andwhen the nuclear installation is in a hot condition:a subset of the plurality of springs consisting of fewer than all of the plurality of springs being in another compressed state and each being compressively engaged with the upper core plate, anda spring of the plurality of springs being in a free state wherein the spring is uncompressed and is disengaged from the upper core plate. 9. The spring apparatus of claim 8 wherein another spring of the plurality of springs has a tail, and wherein the spring has an opening formed therein, the tail being movably received in the opening. 10. The spring apparatus of claim 9 wherein the spring in the free state is disengaged from the fuel assembly. 11. The spring apparatus of claim 8 wherein the spring is situated relatively closer to the nozzle than the other springs of the plurality of springs. 12. The spring apparatus of claim 8 wherein a portion of the spring in the free state is spaced from the other springs of the plurality of springs. 13. The spring apparatus of claim 8 wherein the spring in the free state is of a substantially unvarying thickness.
056028881	summary	FIELD OF THE INVENTION This invention relates to reducing the corrosion potential of components exposed to high-temperature water. As used herein, the term "high-temperature water" means water having a temperature of about 150.degree. C. or greater or steam. High-temperature water can be found in a variety of known apparatus, such as water deaerators, nuclear reactors, and steam-driven power plants. BACKGROUND OF THE INVENTION Nuclear reactors are used in electric power generation, research and propulsion. A reactor pressure vessel contains the reactor coolant, i.e. water, which removes heat from the nuclear core. Respective piping circuits carry the heated water or steam to the steam generators or turbines and carry circulated water or feedwater back to the vessel. Operating pressures and temperatures for the reactor pressure vessel are about 7 MPa and 288.degree. C. for a boiling water reactor (BWR), and about 15 MPa and 320.degree. C. for a pressurized water reactor (PWR). The materials used in both BWRs and PWRs must withstand various loading, environmental and radiation conditions. Some of the materials exposed to high-temperature water include carbon steel, alloy steel, stainless steel, and nickel-based, cobalt-based and zirconium-based alloys. Despite careful selection and treatment of these materials for use in water reactors, corrosion occurs on the materials exposed to the high-temperature water. Such corrosion contributes to a variety of problems, e.g., stress corrosion cracking, crevice corrosion, erosion corrosion, sticking of pressure relief valves and buildup of the gamma radiation-emitting Co-60 isotope. Stress corrosion cracking (SCC) is a known phenomenon occurring in reactor components, such as structural members, piping, fasteners, and welds, exposed to high-temperature water. As used herein, SCC refers to cracking propagated by static or dynamic tensile stressing in combination with corrosion at the crack tip. The reactor components are subject to a variety of stresses associated with, e.g., differences in thermal expansion, the operating pressure needed for the containment of the reactor cooling water, and other sources such as residual stress from welding, cold working and other asymmetric metal treatments. In addition, water chemistry, welding, crevice geometry, heat treatment, and radiation can increase the susceptibility of metal in a component to SCC. It is well known that SCC occurs at higher rates when oxygen is present in the reactor water in concentrations of about 1 to 5 ppb or greater. SCC is further increased in a high radiation flux where oxidizing species, such as oxygen, hydrogen peroxide, and short-lived radicals, are produced from radiolytic decomposition of the reactor water. Such oxidizing species increase the electrochemical corrosion potential (ECP) of metals. Electrochemical corrosion is caused by a flow of electrons from anodic to cathodic areas on metallic surfaces. The ECP is a measure of the kinetic tendency for corrosion phenomena to occur, and is a fundamental parameter in determining rates of, e.g., SCC, corrosion fatigue, corrosion film thickening, and general corrosion. In a BWR, the radiolysis of the primary water coolant in the reactor core causes the net decomposition of a small fraction of the water to the chemical products H.sub.2, H.sub.2 O.sub.2, O.sub.2 and oxidizing and reducing radicals. For steady-state operating conditions, equilibrium concentrations of O.sub.2, H.sub.2 O.sub.2, and H.sub.2 are established in both the water which is recirculated and the steam going to the turbine. This concentration of O.sub.2, H.sub.2 O.sub.2, and H.sub.2 is oxidizing and results in conditions that can promote intergranular stress corrosion cracking (IGSCC) of susceptible materials of construction. One method employed to mitigate IGSCC of susceptible material is the application of hydrogen water chemistry (HWC), whereby the oxidizing nature of the BWR environment is modified to a more reducing condition. This effect is achieved by adding dissolved hydrogen to the reactor feedwater. When the hydrogen reaches the reactor vessel, it reacts with the radiolytically formed oxidizing species on metal surfaces to reform water, thereby lowering the concentration of dissolved oxidizing species in the water in the vicinity of metal surfaces. The rate of these recombination reactions is dependent on local radiation fields, water flow rates and other variables. The injected hydrogen reduces the level of oxidizing species in the water, such as dissolved oxygen, and as a result lowers the ECP of metals in the water. However, factors such as variations in water flow rates and the time or intensity of exposure to neutron or gamma radiation result in the production of oxidizing species at different levels in different reactors. Thus, varying amounts of hydrogen have been required to reduce the level of oxidizing species sufficiently to maintain the ECP below a critical potential required for protection from IGSCC in high-temperature water. As used herein, the term "critical potential" means a corrosion potential at or below a range of values of about -230 to -300 mV based on the standard hydrogen electrode (SHE) scale. IGSCC proceeds at an accelerated rate in systems in which the ECP is above the critical potential, and at a substantially lower or zero rate in systems in which the ECP is below the critical potential. Water containing oxidizing species such as oxygen increases the ECP of metals exposed to the water above the critical potential, whereas water with little or no oxidizing species present results in an ECP below the critical potential. It has been shown that IGSCC of Type 304 stainless steel (containing 18-20% Cr, 8-10.5% Ni, 2% Mn, remainder Fe) used in BWRs can be mitigated by reducing the ECP of the stainless steel to values below -230 mV(SHE). An effective method of achieving this objective is to use HWC. However, high hydrogen additions, e.g., of about 200 ppb or greater, that may be required to reduce the ECP below the critical potential, can result in a higher radiation level in the steam-driven turbine section from incorporation of the short-lived N-16 species in the steam. For most BWRs, the amount of hydrogen addition required to provide mitigation of IGSCC of pressure vessel internal components results in an increase in the main steam line radiation monitor by a factor of five to eight. This increase in main steam line radiation can cause high, even unacceptable, environmental dose rates that can require expensive investments in shielding and radiation exposure control. Thus, recent investigations have focused on using minimum levels of hydrogen to achieve the benefits of HWC with minimum increase in the main steam radiation dose rates. An effective approach to achieve this goal is to either coat or alloy the alloy surface with palladium or other noble metal. Palladium doping has been shown to be effective in mitigating the crack growth rate in Type 304 stainless steel, Alloy 182 having the composition in wt.%: Ni, 59.0 min.; Cr, 13.0-17.0; Fe, 10.0 max.; Mn, 5.0-9.5; Si, 1.0 max.; Cu, 0.5 max.; Ti, 1.0 max.; S, 0.015 max.; C, 0.10 max.; P, 0.03 max.; (Nb+Ta), 1.0-2.5; other, 0.5 max., and Alloy 600 having the nominal composition in wt.%: Cr, 16.0; Fe, 8.0; Si, 0.5; Cu, 0.5 max.; Ti, 0.3 max.; C, 0.08; Ni, balance. The techniques used to date for palladium coating include electroplating, electroless plating, hyper-velocity oxy-fuel, plasma deposition and related high-vacuum techniques. Palladium alloying has been carried out using standard alloy preparation techniques. These approaches are ex-situ techniques in that they cannot be practiced while the reactor is in operation. Also noble metal coatings such as those applied by plasma spraying and by hyper-velocity oxy-fuel must be applied to all surfaces that require protection, i.e., they afford no protection to adjacent uncoated regions. The most critical requirement for IGSCC protection of Type 304 stainless steel is to lower its ECP to values below the protection potential, i.e., -230 mV(SHE). The manner in which this potential is achieved is immaterial, e.g., by alloying, doping or by any other method. It has been demonstrated that it is sufficient to dope the oxide film by the appropriate material (e.g., Pd) to achieve a state of lower ECP in the presence of low levels of hydrogen. It was shown in later work that a thickness of 200-300 .ANG. of the doping element (Pd) is sufficient to impart this benefit of lower potential. This is not surprising because the ECP is an interfacial property, and hence modifying the interface by a process such as doping would alter its ECP. The critical requirement is that the dopant remain on the surface over a long period of time to gain the maximum benefit from the doping action. One method of in-situ application of a noble metal onto stainless steel or other metal surfaces inside a boiling water reactor is by injecting a decomposable noble metal compound into the high-temperature (i.e., 550.degree. F.) water that is in contact with the metal surface during reactor operation. As a result of decomposition of the noble metal compound, the oxide film on the metal surfaces becomes doped with noble metal. The amount of noble metal dopant can be made high enough to provide sufficient catalytic activity for H.sub.2 and O.sub.2 recombination to reduce the ECP of the metal surfaces to required protection values. This approach of noble metal doping has been shown to be effective against crack initiation and crack growth in stainless steel at H.sub.2 /O.sub.2 molar ratios greater than 2 in the reactor environment. SUMMARY OF THE INVENTION The present invention is an alternative method for the application of palladium or other noble metal onto stainless steel or other metal surfaces inside a boiling water reactor. The method comprises the step of injecting a solution of a compound containing a noble metal into the coolant water while the reactor is shutdown or during heatup with only recirculation pump heat, i.e., without nuclear heat generation. As used herein, the term "solution" refers to both solutions and suspensions of a noble metal compound. During shutdown, the reactor coolant water reaches temperatures as low as 120.degree. F., in contrast to the water temperature of 550.degree. F. during normal operation. On the other hand, pump heat can bring the water temperature up to 400.degree.-450.degree. F. At these reduced temperatures, the rate of thermal decomposition of the injected noble metal compound is reduced. However, decomposition of the noble metal compound is also induced by radiation produced inside the reactor. In particular, the noble metal compound can be decomposed by the gamma radiation emanating from the nuclear fuel core of the reactor. Decomposition can also be precipitated by radiation from radioactive isotopes in the material of the reactor component to be protected by noble metal doping. Thus the noble metal compound decomposes under reactor thermal and radiation conditions to release atoms of the noble metal which incorporate in or deposit on the oxide film formed on the stainless steel and other alloy components. As used herein, the term "atoms" also includes ions of the noble metal. The preferred compound for use in noble metal doping is palladium acetylacetonate. The concentration of palladium in the reactor water is preferably in the range of 5 to 100 ppb. Upon injection, the palladium acetylacetonate decomposes and deposits palladium on the crudded (heavily oxidized) stainless steel and other alloy surfaces immersed in the water. The palladium is incorporated into the oxide film or crud via a process wherein palladium ions/atoms apparently replace iron, nickel and/or chromium atoms in the oxide film or crud, resulting in palladium doping. Alternatively, palladium may be deposited within or on the surface of the oxide film or crud in the form of a finely divided metal. The oxide film is believed to include mixed nickel, iron and chromium oxides. The passive oxide films on the surfaces of structural materials can be doped or coated with palladium or other noble metal using either in situ or ex situ techniques. In accordance with both techniques, the structural material is immersed in a solution or suspension of a compound containing the noble metal. Decomposition of the noble metal compound in the solution is then induced by exposing the immersed material to electromagnetic radiation, e.g., ultraviolet or gamma radiation. Even at the low temperatures reached during reactor shutdown or during heatup with recirculation pump heat, the radiation-induced decomposition in combination with thermal decomposition provide noble metal doping of oxide film or crud sufficient to reduce the ECP at the oxide film/water interface to below the critical threshold potential, thereby mitigating stress corrosion cracking.
	abstract	A method for separating amorphous iron oxides is provided. The method includes steps of sampling, filtering, dissolving and separating, analyzing the solution containing amorphous radioactive iron oxides and analyzing granules containing crystalline radioactive iron oxides. Characteristics of the radioactive iron oxides during various periods are acquired to solve the radiation buildup problem. Parameters for improving water quality and chemistry performance indicator are thus provided. Crystalline deposits are separated while the dissolving rate of radioactive iron oxides reaches more than 90%. The present invention does not use complex utilities, is easy to use and has a low operation cost for fast analysis.
039322173	claims	1. A method of passive protection of a nuclear reactor and especially a fast reactor, the steps of replacing at least part of the conventional fuel elements within the core of said reactor by a number of vertical safety fuel elements, forming each safety fuel element with an axial flow duct having a cross-section of large diameter having a ratio with respect to the diameter of the fuel element on the order of 10/15, receiving in said flow duct the portion of fuel which is capable of melting within the central portion of said element under the action of a neutron-flux excursion and flowing the melted fuel portion under gravity from the mid-height of said element to the base thereof thus reducing the reactivity without thereby impairing the fuel can. 2. A device for the passive protection of a nuclear reactor having vertical fuel elements including safety fuel elements wherein each safety element is constituted by a stack of fissile fuel pellets enclosed in a can, each pellet having a central orifice of large diameter having a ratio with respect to the diameter of the fuel element on the order of 10/15 to ensure that the portion of molten fuel melting within the central portion of the element can flow under gravity to the bottom portion of said fuel element. 3. A device according to claim 2, wherein each safety element is provided at the lower end thereof with a member corresponding to the lower fertile blanket of the reactor core, said member being formed of refractory material and provided over part of its height and at the top portion thereof with a central duct having substantially the same diameter as the duct formed in the fissile portion of said fuel element. 4. A device according to claim 3, wherein the member corresponding to the lower blanket is provided at the top portion thereof with a central duct having a diameter which is substantially equal to that of the fissile portion of the fuel element and wherein said member terminates at the lower end thereof in a solid pellet of refractory material. 5. A device for the passive protection of a nuclear reactor having vertical fuel elements including safety fuel elements wherein each safety element is constituted by a stack of fissile fuel pellets enclosed in a can, each pellet having a central orifice of large diameter having a ratio with respect to the diameter of the fuel element on the order of 10/15 to ensure that the portion of molten fuel melting within the central portion of the element can flow under gravity to the bottom portion of said fuel element, a member for each safety element at the lower end thereof corresponding to the lower fertile blanket of the reactor core, said member being formed of refractory material and provided over part of its height and at the top portion thereof with a central duct having substantially the same diameter as the duct formed in the fissile portion of said fuel element; said member having at the bottom portion thereof an axial duct of smaller diameter and beneath said member a chamber for the fission gases which is limited by the bottom face of said member and by the can and at the lower end thereof by a crucible of refractory material on the bottom end-wall of said can. 6. A device according to claim 5, wherein the refractory material is an oxide of fertile substance selected from the group comprising depleted uranium, natural uranium and thorium. 7. A device according to claim 5, wherein the refractory material which forms the bottom crucible is selected from the group comprising depleted uranium, natural uranium, thorium and boron carbide. 8. A device according to claim 7, wherein the safety elements are grouped together in a predetermined number of fuel assemblies of the reactor core. 9. A device according to claim 7, wherein the safety elements are uniformly distributed within all the fuel assemblies of the reactor core. 10. A device according to claim 7, wherein all the fuel elements of the reactor core are safety elements.
	claims	1. A nozzle seal configured to hermetically seal a nozzle that penetrates a reactor vessel wall, the nozzle having a nozzle axis and including a coaxial rod member inserted there through, and a flange portion projecting outward from the nozzle in a radial direction at an end of the nozzle distal from an outer side of the reactor vessel, the nozzle seal comprising:a blocking member that is a plate having a thickness direction parallel to the nozzle axis and that is adjacent to the flange portion along the nozzle axis, the blocking member having a center portion through which the rod member passes;a first seal member provided between the flange portion and the blocking member;a second seal member provided between the rod member and the blocking member;a first fastening portion configured to fasten the flange portion and the blocking member; anda second fastening portion that is adjacent to the first fastening portion configured to fasten the rod member and the blocking member,the blocking member comprising:a first fastened surface configured to receive first fastening force applied by the first fastening portion; anda second fastened surface configured to receive second fastening force applied by the second fastening portion,wherein the first fastened surface and the second fastened surface are continuous and formed on a flat surface of the plate. 2. The nozzle seal according to claim 1, whereinthe blocking member iswhere the rod member coaxially passes through the center of the blocking member, andthe first fastening portion includes a plurality of first hydraulic lock mechanisms arranged around the blocking member, the first hydraulic lock mechanisms being configured to apply the first fastening force by hydraulic pressure so that pressure is uniformly applied along a circumference of the flange portion, the first hydraulic lock mechanisms each comprising a first cylinder pressing the blocking member and a nut applying the first fastening force to the flange portion, and being configured to clamp the blocking member and the flange portion in the thickness direction between the first cylinder and the nut. 3. The nozzle seal according to claim 2, wherein the plurality of first hydraulic lock mechanisms are arranged at predetermined intervals along a circumferential direction of each of the flange portion and the blocking member, and when hydraulic pressure of a same level is collectively applied to the plurality of first hydraulic lock mechanisms, the first hydraulic lock mechanisms apply the first fastening force uniform to each other. 4. The nozzle seal according to claim 1, whereinthe blocking member is a disk where the rod member coaxially passes through the center of the blocking member, andthe second fastening portion includes a second hydraulic lock mechanism arranged in the center of the blocking member, the second hydraulic lock mechanism being configured to apply the second fastening force by hydraulic pressure so that pressure is uniformly applied along a circumference of the rod member, the second hydraulic lock mechanism comprising a second cylinder pressing the blocking member, and being configured to clamp the second seal member in the thickness direction between the second cylinder and the rod member. 5. The nozzle seal according to claim 4, whereinthe second fastening portion further includes a locking member configured to lock the second hydraulic lock mechanism and the rod member.
	claims	1. A method of evaluating a resolution of a scanning electron microscope, comprising the steps of:picking up a first image of a concave and convex pattern formed on a surface of a sample utilizing a first scanning electron microscope;picking up a second image of the concave and convex pattern on the sample utilizing a second scanning electron microscope;respectively processing the first image and the second image in order to evaluate unevenness in resolution between the first scanning electron microscope and the second scanning electron microscope; anddetermining whether a height of the concave and convex pattern as measured from a bottom thereof is sufficient so that no affection by a secondary electron emitted from the bottom of the concave and convex pattern is exhibited. 2. A method of evaluating a resolution of a scanning electron microscope according to claim 1, wherein the concave and convex pattern has a rounded portion where a concave part and a convex part of the concave and convex pattern join. 3. A method of evaluating a resolution of a scanning electron microscope according to claim 1, wherein the first image and the second image of the concave and convex pattern are those which do not substantially include data as to a side wall of the concave and convex pattern. 4. A method of evaluating a resolution of a scanning electron microscope as set forth in claim 1, wherein the first image obtained by picking up the concave and convex pattern with the use of the first scanning electron microscope and the second image obtained by picking up the concave and convex pattern with the used of the second scanning electron microscope, are each obtained by irradiating and scanning an electron beam which is converged so as to have a beam diverging angle smaller than an inclined angle of a side surface of the concave and convex pattern, in the vicinity of a beam waist thereof. 5. A method of evaluating a resolution of a scanning electron microscope as set forth in claim 1, wherein the concave and convex pattern has a backward tapered sectional shape. 6. A method of evaluating a resolution of a scanning electron microscope, comprising the steps of:picking up a first image of a concave and convex pattern formed on a surface of a sample utilizing a first scanning electron microscope;picking up a second image of the concave and convex pattern on the sample with the use of a second scanning electron microscope;respectively processing the first image and the second image in order to evaluate unevenness in resolution between the first scanning electron microscope and the second scanning electron microscope; anddetermining whether a top of the concave and convex pattern measured from a bottom thereof is enough high so that a secondary electron emitted from the bottom of the concave and convex pattern cannot reach a first secondary electron detector of the first scanning electron microscope or a second secondary electron detector of the second scanning electron microscope. 7. A method of evaluating a resolution of a scanning electron microscope according to claim 6, wherein the concave and convex pattern has a rounded portion where a concave part and a convex part of the concave and convex pattern join. 8. A method of evaluating a resolution of a scanning electron microscope according to claim 6, wherein the first image and the second image of the concave and convex pattern are those which do not substantially include data as to a side wall of the concave and convex pattern. 9. A method of evaluating a resolution of a scanning electron microscope as set forth in claim 6, wherein the first image obtained by picking up the concave and convex pattern with the use of the first scanning electron microscope and the second image obtained by picking up the concave and convex pattern with the used of the second scanning electron microscope, are each obtained by irradiating and scanning an electron beam which is converged so as to have a beam diverging angle smaller than an inclined angle of a side surface of the concave and convex pattern, in the vicinity of a beam waist thereof. 10. A method of evaluating a resolution of a scanning electron microscope as set forth in claim 6, wherein the concave and convex pattern has a backward tapered sectional shape.
052767183	abstract	In a field of commercial nuclear reactors, a control blade, which is inserted into and withdrawn from the core of a nuclear reactor so as to start/shut-down the operation thereof and to control the reactor power, must have soundness, high reactivity and long life. The control blade for nuclear reactors is arranged to prevent swelling of a neutron absorber so as to improve the soundness. Furthermore, even if the swelling takes place, the soundness of the control blade can be maintained. The control blade is arranged to prevent the swelling phenomenon in such a manner that boron exhibiting excellent neutron absorbing performance absorbs hydrogen which is produced as a result of reactions with the neutrons. The control blade for nuclear reactors has an upper structure member, a lower structure member, a central tie member disposed between the upper and lower structure members, a plurality of wings connected to each other by the central tie member, and neutron absorber enclosed in the wing. Each of the wings is constituted by a plate member made of hafnium metal, a hafnium alloy or an alloy the main component of which is zirconium or titanium. The neutron absorber comprises a long-lived type neutron absorber which is enclosed in accommodating holes formed in the front insertion portion of the wing which is exposed to a large amount of neutrons and a neutron absorber which is inserted into at least the major portion of the residual accommodating holes and which contains boron. A mixture of a material containing boron and a hydrogen absorber is enclosed in the accommodating holes for accommodating the neutron absorber containing boron disposed in a range from the front insertion portion, which is exposed to a large amount of neutrons, to 1/4 of the height L of an effective core portion of a nuclear reactor.
	summary
	claims	1. A final, ready to use, spacer grid configured to separate and hold nuclear fuel rods in a nuclear reactor of the boiling water reactor type in predetermined positions relative to each other, wherein the final spacer grid comprises:i) a spacer grid structure made of an alloy that has been formed and assembled such that it constitutes a spacer grid, andii) an outer oxide coating on the surface of the spacer grid structure, said outer oxide coating comprises a first inner oxide layer of a first composition and a second outer oxide layer of a second composition different from the first composition, wherein the first inner oxide layer mainly contains Cr2O3 and the second outer oxide layer mainly contains NiFe2O4;wherein said alloy is a Ni base alloy that consists of the following:Element% by weightNi>50.0Cr14.0-21.0Fe12.0-23.0Ti1.50-3.0 Al0.40-1.50Co0.0001-0.010 C0.001-0.050N0.001-0.030Nb0.001-1.50 Ta0.001-0.030Si0.01-0.50Mn0.01-1.0 S0.001-0.020P0.001-0.050Cu0.01-0.50Mo + W0.001-1.0 the total amount of one or more elements chosen from the 0-1.0group consisting of all elements except for the elementsreferred to in the table above. 2. A final spacer grid according to claim 1, wherein the amount of Fe in said alloy is 15.0-19.0% by weight. 3. A final spacer grid according to claim 1, wherein the amount of Co in said alloy is <0.0050% by weight. 4. A final spacer grid according to claim 1, wherein said alloy in the final spacer grid comprises a substantial amount of γ′ secondary phase particles such that the final spacer grid has a sufficient mechanical strength. 5. A final spacer grid according to claim 4, wherein the mole fraction of γ′ secondary phase particles in said alloy in the final spacer grid is 5-25%. 6. A final spacer grid according to claim 1, wherein said outer oxide coating has a thickness of 50-1000 nm. 7. A final spacer grid according to claim 1, wherein the first inner oxide layer has a thickness of 50-200 nm and the second outer oxide layer has a thickness of 20-80 nm. 8. A final spacer grid according to claim 1, wherein said alloy is a Ni base alloy that consists of the following:Element% by weightNi>60.0Cr14.0-17.0Fe15.0-19.0Ti1.750-2.750Al0.40-1.0 Co0.0001-0.0050C0.001-0.050N0.001-0.030Nb0.70-1.20Ta0.001-0.030Si0.01-0.50Mn0.01-1.0 S0.001-0.010P0.001-0.020Cu0.01-0.50Mo + W0.001-0.20 the total amount of one or more elements chosen from the 0-0.50group consisting of all elements except for the elementsreferred to in the table above. 9. A method of manufacturing the final, ready to use, spacer grid comprising the steps of:providing a Ni base alloy that consists of the following:Element% by weightNi>50.0Cr14.0-21.0Fe12.0-23.0Ti1.50-3.0 Al0.50-1.50Co0.0001-0.010 C0.001-0.050N0.001-0.030Nb0.001-1.50 Ta0.001-0.030Si0.01-0.50Mn0.01-1.0 S0.001-0.020P0.001-0.050Cu0.01-0.50Mo + W0.001-1.0 the total amount of one or more elements chosen from the 0-1.0group consisting of all elements except for the elementsreferred to in the table above.forming and assembling the alloy such that said spacer grid structure is obtained, andheat treating the spacer grid structure at a temperature of 650-750° C. for 5-23 hours, the heat treatment being performed in an oxidizing atmosphere, wherein the heat treatment is such that an outer oxide coating is formed on the surface of the spacer grid structure comprising a first inner oxide layer of a first composition and a second outer oxide layer of a second composition different from the first composition such that the first inner oxide layer mainly contains Cr2O3 and the second outer oxide layer mainly contains NiFe2O4, thereby obtaining the final, ready to use, spacer grid. 10. A method of manufacturing the final spacer grid according claim 9, wherein said oxidizing atmosphere comprises aqueous vapour and air. 11. A method of manufacturing the final spacer grid according to claim 9, wherein said heat treatment of the spacer grid structure is such that γ′ secondary phase particles are formed in said alloy, thereby obtaining improved mechanical properties of the final spacer grid. 12. A method of manufacturing the final spacer grid according to claim 11, wherein the mole fraction of γ′ secondary phase particles in said alloy in the final spacer grid is 5-25%. 13. A method of manufacturing the final spacer grid according to claim 9, wherein the first inner oxide layer has a thickness of 50-200 nm and the second outer oxide layer has a thickness of 20-80 nm.
	abstract	A charged particle beam irradiation device includes an accelerator that accelerates charged particles and emits a charged particle beam; an irradiation unit that irradiates a body with the charged particle beam; a duct that transports the charged particle beam to the irradiation unit; a tubular body arranged on a propagation path of the charged particle beam within the irradiation unit, has inert gas filled thereinto, and has particle beam transmission films transmitting the charged particle beam therethrough at an inlet and an outlet thereof; a gas supply unit that supplies the inert gas into the tubular body; and a leak valve that leaks the inert gas inside the tubular body to the outside when the internal pressure of the tubular body is equal to or higher than a set pressure. The gas supply unit has a plurality of supply lines having different amounts of supply of inert gas.
043057861	description	DESCRIPTION OF THE PREFERRED EMBODIMENT The drawings show an illustrative embodiment of the invention in a storage rack for spent nuclear fuel assemblies. The rack 10 is shown generally in FIG. 1 as comprising a plurality of identical rectangular storage cells 12 which can be aligned with each other in a regular or checkerboard array. The rack 10 is shown alone in FIG. 1 but it will be understood that it is intended to be immersed in a water-filled pit or pool or otherwise filled with water that may or may not contain other chemicals. The cells 12 are all identical and preferably square in cross section and are made of sufficient length to contain the fuel assemblies 14 to be stored. The cells, which are formed from partitions 16, are arranged in a regular array, and are preferably secured together by welding or in any other desired manner to form a modular structure. The cells 12 are preferably made of stainless steel which is a satisfactory structural material and which has the ability to absorb neutrons so that it also serves as a poison material. Any suitable material including those containing deliberate neutron absorbing elements may be used, however, for making the cells. In carrying out the invention, a neutron source may be placed somewhere in the region of a storage rack, such as that of FIG. 1, where the nuclear fuel will eventually be located or the fuel elements themselves may be considered to be the neutron sources. A neutron detector or a series of detectors are placed inside or outside of the storage tank; and the counting rate of neutrons as they arrive at the detectors is observed. The nuclear fuel assemblies are then added to the region one at a time and the counting rate is observed as the fissionable material builds up the multiplication factor of the storage rack. The counting rate detected by the neutron detector follows roughly the well-known subcritical multiplication formula: EQU CR=.alpha..sub.S /(1-k) (1) where: S is the neutron source emission rate; PA1 .alpha. is a term that reflects sensitivity of the detector as well as the attenuation of the neutrons between source and detector, and various geometric considerations to be described later; PA1 CR is the counting rate from the neutron detector and associated electronics; and PA1 k, as before, is the multiplication factor of the assembly at any given time for any given element configuration. A plot, such as the plot of FIG. 2, is usually made of the inverse counting rate, 1/CR, as a function of the number of fuel elements or fuel assemblies added to the storage tank. From this plot, it will be appreciated that the critical point will occur when 1/CR approaches zero. It will be noted from FIG. 2 that for low multiplication factors (i.e., high values of 1/CR), the points appear to be somewhat random. Meaningful predictive information from this plot is not available until high multiplication factors of k greater than approximately 0.9 have been reached. (See Equation (1) above.) Before describing the invention in detail, and by way of background, it will first be assumed that a storage tank for spent fuel assemblies contains a neutron source and a detector, connected to the necessary electronics to form a count rate meter; and that fuel assemblies are being added to the pool one by one. (See FIG. 3.) If a reading is taken on the count rate meter connected to the detector when the multiplication factor is very low (multiplication factor, k, is unknown, but for illustrative purposes is assumed to be 0.1). Then: EQU CR.sub.1 =.alpha..sub.1 S/(1-k.sub.1) (2) where the "1" subscripts refer to this initial reading. Note that if k.sub.1 were zero, .alpha..sub.1 S would be divided by 1, and if k.sub.1 were 0.1,.alpha..sub.1 S would be divided by 0.9, not a large difference in counting rate. The significance of this point is that the first counting rate reading is not particularly sensitive to k.sub.1. Let it be further assumed that more and more fuel assemblies are now loaded into the tank or pool in some fixed geometric pattern, and assume further that it is desirable to stop the loading process at, say, k=0.95. The reading of the counting rate meter can then be represented as: EQU CR.sub.2 =.alpha..sub.2 S/(1-k.sub.2) (3) where the "2" subscripts refer to the second reading after loading is stopped. The ratio of CR.sub.2 /CR.sub.1 can be represented as: EQU CR.sub.2 /CR.sub.1 =(.alpha..sub.2 /.alpha..sub.1).(1-k.sub.1)/(1-k.sub.2) (4) In accordance with the present invention, the value of k.sub.2 is calculated from the foregoing equation; and if k.sub.2 is too close to unity, the loading process is stopped or other modifications are made since criticality is being approached. Assuming that CR.sub.2, CR.sub.1, and k.sub.1 are known, k.sub.2 can be computed from the foregoing equation simply if it is assumed that .alpha..sub.2 /.alpha..sub.1 is unity. As a practical matter, however, .alpha..sub.2 /.alpha..sub.1 is not unity and can vary and depends upon various geometric considerations. There are three principal reasons why the ratio of .alpha..sub.2 to .alpha..sub.1 is not unity. To begin with, the neutron source-to-detector attenuation will vary. This can be illustrated, for example, with the use of FIG. 3 where the numeral 20 designates a storage rack for nuclear fuel assemblies containing a plurality of cells numbered 1 through 14 for the respective fuel assemblies. The neutron source in FIG. 3 is identified by reference numeral 22; while the neutron detector is identified by the reference numeral 24. It will be assumed that the source 22 and detector 24 are initially in place with only water and structural material between them. A fuel assembly is then placed in cell No. 1 in the pool, followed by placing assemblies in cell Nos. 2, 3, 4, etc. From FIG. 3, it is apparent that the neutrons coming from the source must be attenuated differently by each addition of fuel material irrespective of the effective neutron multiplication. The ratio of .alpha..sub.2 to .alpha..sub.1 also varies by virtue of a change in neutron flux distribution as illustrated in FIG. 4. With specific reference to FIG. 4, when a source is placed in the center of a low multiplication region, a neutron flux distribution results in the material that is sharply peaked in the neighborhood of the source and is indicated by curve 26 in FIG. 4. As more and more fissionable material is added to the assembly, the flux distribution changes its shape as indicated by curves 28 and 30 of FIG. 4. Curve 26 results from a low multiplication factor, k, curve 28 results from a higher multiplication factor, and curve 30 results when k is near unity. As the multiplication factor closely approaches unity, the flux distribution shape approaches that of an operating critical reactor. These changes in flux shape affect the reading of a count rate meter whose detector is placed in the vicinity of the assembly. Finally, the ratio of .alpha..sub.2 to .alpha..sub.1 can vary due to changes in detector reading as a function of the loading sequence for nuclear reactor fuel assemblies. Consider, for example, the situation of FIG. 5 where a source 22 within a spent fuel pit 20 is initially symmetrically surrounded by a few fuel elements. The detector 24, for convenience, is shown as being displaced from the pit 20. If it is now desired to add the next element (shown dotted in position 1 or 2 in FIG. 5), the addition of the next element in either position 1 or 2 increases the multiplication factor, k of the assembly by the same amount. Yet, it is obvious that the neutron detector 24 will read differently depending upon whether position 1 or position 2 is loaded next. The physical effect is a combination of the problems previously mentioned; but for ease of correction all effects may be considered to be a function of loading pattern. It will also be apparent that if one were to start building up a fissionable mass in one end of the pool and another fissionable mass in the opposite end of the pool, at some point the two masses will begin to interact and each add to the multiplication factor of the other. Some form of coupling can exist in some types of large reactors and care must be taken that multiple reactors do not exist in the same pool. In accordance with the present invention, the solution for all of these problems begins first by having a rigid loading sequence and pattern for the insertion of spent fuel assemblies into the storage pool. In this manner, the pool is always loaded in exactly the same way. The counting rate, CR, may now be considered as consisting of two parts: A counting rate that is dependent on the amount of fissionable material and poison present, and correction factor to the counting rate, .alpha..sub.2 /.alpha..sub.1, that is essentially only positionally dependent. As each successive element is loaded, the actual counting rate is read, and this information, plus the position of the successive element in the pool, is fed into a small microcomputer or minicomputer associated with the counting rate meter. The computer contains a memory in which is stored the correction factor, .alpha..sub.2 /.alpha..sub.1 for each pool position. The computer then proceeds to calculate and display the multiplication factor, k.sub.2, based on the equation: EQU CR.sub.2 /CR.sub.1 =(.alpha..sub.2 /.alpha..sub.1).(1-k.sub.1)/(1-k.sub.2) To make the correction factors smaller, the first multiplication factor, k, is not measured until a configuration similar to FIG. 3 is reached. In this way, drastic changes in source neutron attenuation are avoided. The correction factors, .alpha..sub.2 /.alpha..sub.1, for each pool position are determined empirically or can be obtained from mathematical considerations with the use of a large computer and sophisticated codes that are well known in the art. It will be noted from FIG. 2, for example, that as k approaches unity, the accuracy of the measurement becomes better, and depends more on the basic multipication factor and less on the correction factor. Alarms and safety devices can be triggered when the calculated counting rate ratio used in the measurement is about twenty, which represents a large safety factor in not allowing a critical loading. The multiple reactor coupling problem can be handled by multiple sources and detectors if the problem is shown to exist theoretically for a given pool design. Such an arrangement, for example, is shown in FIG. 6 where there are three neutron sources 32, 34 and 36 within a fuel storage pool 40 and four neutron detectors 42A, 42B, 42C and 42D. The detectors, in turn, are connected to four count rate meters 44A-44D which produce output electrical signals, each proportional to its associated count rate, which are fed to a small computer 46. This small computer includes a memory wherein the correction factors .alpha..sub.2 /.alpha..sub.1 are stored for each fuel assembly with a given geometrical configuration based upon the assumption that the pool is always loaded in exactly the same way. Also stored in the computer 46 are the count rates obtained after the initial fuel assemblies were inserted into the pool. From these factors, the value of k for each fuel assembly loaded into the pool can be displayed on a display 48; and if k should closely approach unity, an alarm 50 can be actuated. It should be noted that in an installation such as that shown in FIG. 6, the gamma ray environment of the neutron detectors is extremely hostile. Conventional fission counters have been shown to have long lives even in gamma fields up to 10.sup.5 R/hour. As the fields adjacent to a used fuel element immediately after reactor shutdown can be considerably higher than this amount, lead shielding or the like may be provided for the detectors. An efficient detector package would be one that occupies one fuel assembly cell. If this amount of shielding material were shown to be insufficient, however, it may be necessary to restrict moving the elements into the pool until a day or two after reactor shutdown. The sources, such as sources 32-36 shown in FIG. 6 do not have the same problem as the detectors in that they are small, can withstand the environment, and can be placed in suitable positions in the fuel assembly crate lattice. Although the invention has been shown in connection with certain specific embodiments, it will be readily apparent to those skilled in the art that various changes in form and arrangement of parts may be made to suit requirements without departing from the spirit and scope of the invention.
046718974	summary	BACKGROUND OF THE INVENTION This invention relates to a process and apparatus for solidification of radioactive waste occurring in a nuclear power station, and more particularly to a process and apparatus for its solidification utilizing a hydraulic solidifier. The amounts of radioactive waste occurring in nuclear power stations and related facilities have been increasing year by year, and a need for the volume reduction of such radioactive waste has consequently been increasing in order to secure a storage space within the facilities. Methods which have so far been examined for the volume reduction of radioactive waste includes the following. Concentrated liquid waste obtained by concentrating the liquid waste formed in the regeneration of spent ion-exchange resin and the slurry of powdery ion-exchange resin which occur in large amounts in a nuclear power station are dried into powder. Thus the liquid waste is freed of water which accounts for the major or part of its volume and, if necessary, the powder is further pelletized and solidified collectively by packing in a solidification vessel. (U.S. Pat. No. 4,299,271). However, such a method is still defective in that the liquid waste cannot necessarily be converted into a stable solid when using a hydraulic solidifier such as cement or alkali silicates (e.g. water glass). The concentrated liquid waste occurring in a boiling water reactor (BWR) nuclear power station is composed chiefly of a sodium salt, i.e. sodium sulfate (Na.sub.2 SO.sub.4). In a pressurized water reactor (PWR) nuclear power station, on the other hand, the concentrated liquid waste is composed chiefly of a sodium salt, i.e. sodium borate (Na.sub.2 B.sub.4 O.sub.7). These sodium salts are both water-soluble. In case the concentrated liquid waste occurring in a BWR nuclear power station is dried, powdered or, if necessary, further pelletized, and then solidified with a hydraulic solidifier, sodium sulfate which is its main ingredient will absorb free water contained in the solidifier paste and water formed by the solidification reaction, and thereby form a swollen hydrate, Na.sub.2 SO.sub.4.10H.sub.2 O, to cause cracking in the solidified body. In addition, in case of a cement solidifier, sodium sulfate will react with calcium hydroxide which is formed when cement is hydrated, and thereby form gypsum, which will prevent the cement from hardening too rapidly but will, on the other hand, accelerate the formation of ettringite (3CaO.Al.sub.2 O.sub.3.3CaSO.sub.4.32H.sub.2 O) to cause the solidified body to be swollen or broken. In case the concentrated liquid waste occurring in a PWR nuclear power station is solidified, sodium borate, which is the main ingredient, will likewise cause the solidified body to lower its strength. It will form a hydrage, Na.sub.2 B.sub.4 O.sub.7.10H.sub.2 O, to generate heat. In the case of a cement solidifier, it will inhibit the formation of a hydrate of calcium silicate (3CaO.2SiO.sub.2.3H.sub.2 O) and of a hydrate of calcium aluminate (3CaO.Al.sub.2 O.sub.3.6H.sub.2 O) by the hydration of cement. Since, in either of these cases, the powdered or pelletized waste mainly comprises the water-soluble sodium salts, the solidified body suffers from degradation of its structure, reduction in the leaching rate, and lowering in the strength and specific gravity owing to exudation during a prolonged storage. In the solidification procedure, furthermore, sodium borate reacts with the hydraulic solidifier very promptly, and the solidification proceeds so rapidly as to disturb the smooth pouring of the solidification mixture. To prevent this, the content of the liquid waste in the solidification mixture will have to be limited to at most 30 wt % and the volume reduction ratio be correspondingly lowered. SUMMARY OF THE INVENTION The object of the present invention is to provide a process for the solidification of radioactive waste wherein the solidified body is obtained having high consistency for a long time. Another object of the present invention is to provide a process for the solidification of radioactive waste wherein the solidified body is obtained having high volume reduction. Another object of the present invention is to provide a process for the solidification of radioactive waste wherein the solidified body is obtained having less degradation of its structure owing to exudation. Another object of the present invention is to provide a process for the solidification of radioactive waste wherein the solidified body is obtained having low leaching rate. The inventors have drawn their attention to the finding that the above-mentioned problems are all due to the soluble salt contained as the main ingredient in the liquid waste. Thus, they have made various studies in the belief that these problems could be solved by converting the radioactive waste into a hardly water-insoluble salt structure (including an insoluble structure) before it is submitted to a solidification process, and have finally attained the present invention. The process for solidification of radioactive waste according to this invention is characterized in that the radioactive waste is first converted into a hardly water-soluble powder (including a water-insoluble powder) and then solidified with a hydraulic solidifier in a solidification vessel. The radioactive waste may be powdered (including granulated and encapsulated) by incorporating the radioactive waste with a substance which is capable of reacting with the water-soluble salt contained in said radioactive waste to form a hardly water-soluble salt (including a water-insoluble salt) and then powdering the mixture with drying, or by powdering the radioactive waste with drying, granulating the powder with drying and then microencapsulating the granules with a hardly water-soluble substance (including water-insoluble substance). Na.sub.2 SO.sub.4 and Na.sub.2 B.sub.4 O.sub.7 which are main ingredients of liquid radioactive waste occurring in a nuclear power station have high solubilities in water. In order to see how Na.sub.2 SO.sub.4 could be converted into salts hardly soluble (including insoluble) in water and what type of salts they should be, and with attention drawn to the fact that alkaline earth metal sulfates and metal chelate salts were hardly soluble in water, in general, the present inventors selected calcium sulfate, strontium sulfate and barium sulfate for the former, and ammonium cobalt oxalate sulfate and hexaammonium chromium sulfate for the latter to examine their solubilities. The results are shown in Table 1. This table shows the values observed at 20.degree. C. TABLE 1 ______________________________________ Solubility Solubility Substance (wt %) Substance (wt %) ______________________________________ sodium sulfate 16.0 ammonium cobalt 0.8 calcium sulfate 0.205 oxalate sulfate strontium sulfate 1.3 .times. 10.sup.-2 hexaammonium 2.7 barium sulfate 2 .times. 10.sup.-4 chromium sulfate ______________________________________ It was found that all these substances had lower solubilities than sodium sulfate and that conversion into barium sulfate was more effective than into the rest for the intended purpose. In respect of cost, however, conversion into calcium sulfate was thought to be most economical and most practical. Various borates were also tested for solubilities, and conversion into calcium borate was likewise found to be appropriate in respect of cost and practical application. The radioactive waste materials which can be solidified by the procedures include not only dried granulates of concentrated liquid waste and sludge consisting of sodium sulfate, sodium borate, etc. but also a slurry waste of ion-exchange resin, and the so-called miscellaneous solid matters, such as HEPA filters, vinyl sheet clothings and wooden pieces, and their fragments. The solidifer includes not only an alkali silicate composition but also fluid solidifier, such as a thermosetting or thermo-fusible plastic, asphalt, mortar or cement. According to this invention providing a process and apparatus in which the dry powder obtained from the radio-active waste occurring in a nuclear power station is solidified with a hydraulic solidifier, the solidified body can not only be extensively protected from its deterioration and damage caused by water absorption, hydration, exothermic reaction, swelling and leaching due to the sodium sulfate and sodium borate contained in the radioactive waste to thereby retain its consistency for a long time, but also be improved outstandingly in volume reduction ratio.
	summary
044407173	claims	1. An apparatus for sensing the level of a liquid contained within a pressure vessel comprising: a. a plurality of axially spaced elongated heat conductive housings, each of said housings having b. means for supplying energy to each of said heating elements; c. means for measuring the net voltage generated by each of said first thermocouple junctions and said second thermocouple junctions; d. a first tubular element having an interior, an exterior, an enclosed lower end, and means for fluid communication between said interior and said exterior said first tubular element surrounding said plurality of axially spaced elongated heat conductive housings and extending substantially the length of said housings, whereby liquid or liquid flashing to gas within said first tubular element exits through said fluid communicating means upon a pressure reduction or change in fluid level in said pressure vessel thereby maintaining equalized liquid levels on said interior and exterior of said first tubular element, e. a second tubular element having an interior, an exterior, and means for fluid communication between said interior and said exterior, said second tubular element surrounding said first tubular element, and extending substantially the length of said first tubular element; and f. at least a portion of said fluid communicating means between said interior and said exterior of said second tubular element are axially offset from said fluid communicating means between said interior and said exterior of said first tubular element. means for sensing the absolute temperature of at least one of said first thermocouple junctions. means for sensing the absolute temperature of at least one of said second thermocouple junctions. means for reducing power to at least one of said heaters when the temperature of at least one of corresponding said first thermocouples becomes excessive. 2. An apparatus as in claim 1 wherein: said fluid communicatng means between said interior and said exterior of said first tubular element are lateral input-output ports near the top and bottom thereof. 3. An apparatus as in claim 2 wherein: said fluid communicating means between said volume enclosed by said splash guard and said exterior of said splash guard are lateral inlet-outlet ports near the top and bottom of said splash guard. 4. An apparatus as in claim 1, 2 or 3 also comprising 5. An apparatus as in claim 1, 2, or 3 also comprising: 6. An apparatus as in claim 4 also comprising: 7. An apparatus of claim 1, 2 or 3 wherein: (a) said first tubular element is hydraulically divided forming at least two level sensing sections, each of said level sensing sections having said fluid communicating means between said interior and said exterior of said first tubular element and at least one elongated heat conductive housing, (b) said second tubular element is hydraulically divided proximate said hydraulic division in said first tubular element forming said at least two level sensing sections, each of said level sensing sections having said fluid communicating means between said interior and said exterior of said second tubular element, whereby at least two separate liquid levels may be measured. 8. An apparatus as in claim 7 also comprising: means for sensing the absolute temperature of at least one of said first thermocouple junctions. 9. An apparatus as in claim 7 also comprising: means for sensing the absolute temperature of at least one of said second thermocouple junctions.
	abstract	An apparatus for performing automated in-situ lift-out of a sample from a specimen includes a computer having a memory with computer-readable instructions, a stage for a specimen and a nano-manipulator. The stage and the nano-manipulator are controlled by motion controllers connected to the computer. The nano-manipulator has a probe tip for attachment to samples excised from the specimen. The computer-readable instructions include instructions to cause the stage motion controllers and the nano-manipulator motion controllers, as well as an ion-beam source, to automatically perform in-situ lift-out of a sample from the specimen.
	claims	1. A collimator for a beam detector, comprising:at least two juxtaposed collimator plates; andat least one supporting element, arranged between at least two juxtaposed collimator plates, for stiffening the collimator, the at least one supporting element being constructed from a material transparent to X-rays and supporting the collimator plates laterally. 2. The collimator as claimed in claim 1, wherein the collimator includes at least two supporting elements, interconnected in such a way that the collimator includes a device including slots between the supporting elements, each one slot being provided for holding a collimator plate. 3. The collimator as claimed in claim 2, wherein the collimator includes two devices with slots between the supporting elements for holding collimator plates, in which one device serves as a base element and the other device serves as a cover element. 4. The collimator as claimed in claim 2, wherein the slots are configured in such a way that, in the event of arrangement over a beam detector, the collimator plates arranged in the slots are aligned at least substantially with a focus of a radiation source assigned to the beam detector. 5. The collimator as claimed in claim 1, wherein the supporting elements are designed as supporting crosses. 6. The collimator as claimed in claim 1, wherein the supporting elements are of at least one of U-, V- and W-shaped design. 7. The collimator as claimed in claim 1, wherein the supporting elements include at least substantially the same wall thickness. 8. The collimator as claimed in claim 1, wherein the supporting elements are made from a glass fiber reinforced liquid crystal polymer. 9. The collimator as claimed in claim 2, wherein the device having slots is an injection-molded part. 10. The collimator as claimed in claim 2, wherein the device having slots includes a support strut at least one of on the edge side and arranged between two supporting elements. 11. The collimator as claimed in claim 1, wherein the supporting elements are bonded to the collimator plates. 12. The collimator as claimed in claim 11, wherein the adhesive is a low-viscosity adhesive. 13. The collimator as claimed in claim 3, wherein the base element includes at least one positioning lug for the purpose of positionally accurate arrangement over a beam detector. 14. The collimator as claimed in claim 1, wherein the collimator plates include at least one of tungsten, molybdenum and tantalum. 15. An X-ray detector, comprising the collimator as claimed in claim 1. 16. The collimator as claimed in claim 1, wherein the collimator is alignable on all sides next to collimators of identical design. 17. A computed tomography unit comprising a collimator as claimed in claim 1. 18. The collimator as claimed in claim 3, wherein the slots are configured in such a way that, in the event of arrangement over a beam detector, the collimator plates arranged in the slots are aligned at least substantially with a focus of a radiation source assigned to the beam detector. 19. An X-ray detector, comprising the collimator as claimed in claim 2. 20. A computed tomography unit, comprising a plurality of collimators as claimed in claim 1, each aligned on all sides next to collimators of identical design.
	summary
047284846	claims	1. An apparatus for handling a control rod driving mechanism (CRD), comprising: a turn carriage provided in a pedestal space below a pressure vessel of a nuclear reactor; a truck adapted to run along a rail arranged on said turn carriage; CRD conveyor means arranged for being transferred into and out of said pedestal space with said CRD contained therein from and to a space for at least one of inspection and maintenance of said CRD, said CRD conveyor means including lifter means movable therein in a longitudinal direction of said conveyor means with said CRD mounted thereon so as to shift said CRD to and from a CRD housing disposed under said pressure vessel; holding means for removably holding said conveyor means, said holding means being mounted on said truck and being rotatable between an upright position and a horizontal position in said pedestal space; CRD mounting/demounting means for enabling mounting of said CRD to said CRD housing and for enabling demounting of said CRD from said CRD housing, said mounting/demounting means including means fixedly attached to said holding means and means movable transversely of said holding means onto and from said lifter means, said mounting/demounting means being arranged for vertical shifting by said lifter means; and CRD supporting means arranged on said holding means for temporarily supporting said CRD during movement of said CRD between sid CRD conveyor means and said CRD mounting/demounting means. a turn carriage provided in a pedestal space below a pressure vessel of a nuclear reactor; a truck adapted to run along a rail arranged on said turn carriage; CRD conveyor means arranged for being transferred into and out of said pedestal space with said CRD contained therein from and to a space for at least one of inspection and maintenance of said CRD, said CRD conveyor means including lifter means movable therein in a longitudinal direction of said conveyor means with said CRD mounted thereon so as to shift said CRD to and from a CRD housing disposed under said pressure vessel; holding means for removably holding said conveyor means, said holding means being mounted on said truck and being rotatable between an upright portion and a horizontal position in said pedestal space; CRD mounting/demounting means for enabling mounting of said CRD to said CRD housing and for enabling demounting of said CRD from said CRD housing, said mounting/demounting means including means fixedly attached to said holding means and means movable transversely of said holding means onto and from said lifter means; and CRD supporting means for temporarily supporting said CRD during movement of said CRD between said CRD conveyor means and said CRD mounting/demounting means. 2. An apparatus according to claim 1, wherein said CRD mounting/demounting means includes mounting/demounting header means for enabling fitting and removing of bolts for fixing said CRD to said CRD housing and for enabling coupling and uncoupling of a control rod to and from said CRD. 3. An apparatus according to claim 2, wherein said CRD mounting/demounting means further includes a bearing attached to said holding means, and a lift pipe arranged for up and down movement by said lifter means of said CRD conveyor means while being guided by said bearing, said mounting/demounting header means being secured to an upper end of said lift pipe. 4. An apparatus according to claim 3, wherein said bearing is secured to said holding means through a cylinder so that said CRD mounting/demounting means is moved onto said lifter means of said CRD conveyor means when said bearing is moved towards said CRD conveyor means by the action of said cylinder. 5. An apparatus according to claim 3, wherein said CRD mounting/demounting means enables disposing of a drain from said CRD housing and said CRD. 6. An apparatus according to claim 1, wherein said lifter means of said CRD conveyor means engages a screw rod so that said lifter means is moved up and down in response to rotation of said screw rod. 7. An apparatus according to claim 1, wherein said CRD supporting means is provided on an upper end of said holding means. 8. An apparatus according to claim 7, wherein said CRD supporting means is arranged for movement towards said CRD conveyor means so as to support said CRD in said CRD conveyor means. 9. An apparatus for handling a control rod driving mechanism (CRD), comprising:
	claims	1. A method for monitoring the health of a computer system, comprising:receiving a variance of a time series for a monitored telemetry variable within the computer system,calculating a residual function of the variance of the time series for the monitored telemetry variable;calculating a first-difference function of the variance of the time series from the residual function;determining whether the first-difference function indicates that the computer system is at the onset of degradation; andif so, performing a remedial action. 2. The method of claim 1, wherein prior to receiving the variance for the time series for the monitored telemetry variable, the method further comprises:receiving the time series for the monitored telemetry variable; andcalculating the variance of the time series for the monitored telemetry variable. 3. The method of claim 1, wherein calculating the first-difference function of the time series involves, for each time point within the time series, subtracting a value of the time series at a previous time point from the value of the time series at a present time point. 4. The method of claim 3, further comprises dividing the result of the subtraction by the value of a length of a time interval between the previous time point and the present time point. 5. The method of claim 1, wherein calculating the residual function for a time series involves:for each time interval in the time series,calculating a running average of values for the time series up to and including a present time interval; andsubtracting the running average from a value of the time series at the present time interval. 6. The method of claim 1, wherein determining whether the first-difference function indicates that the computer system is at the onset of degradation involves determining whether the first-difference function exceeds a specified threshold. 7. The method of claim 1, wherein determining whether the first-difference function indicates that the computer system is at the onset of degradation involves:performing a Sequential Probability Ratio Test (SPRT) on the first-difference function; anddetermining whether the SPRT generates an alarm. 8. The method of claim 7, wherein the SPRT can include one or more of:a positive variance first-difference test, which generates an alarm if the first-difference function for the variance of the time series for the monitored telemetry variable is increasing; anda negative variance first-difference test, which generates an alarm if the first-difference function for the variance of the time series for the monitored telemetry variable is decreasing. 9. The method of claim 1, wherein performing the remedial action can involve performing one or more of:recording a time when the onset of degradation occurred;notifying a system administrator that the computer system is at the onset of degradation;shutting down the computer system;backing up data stored on the computer system;failing-over to a redundant computer system; andreplacing one or more components which are at the onset of degradation. 10. A computer-readable storage medium storing instructions that when executed by a computer cause the computer to perform a method for monitoring the health of a computer system, wherein the method comprises:receiving a variance of a time series for a monitored telemetry variable within the computer system;calculating a residual function of the variance of the time series for the monitored telemetry variable;calculating a first-difference function of the variance of the time series from the residual function;determining whether the first-difference function indicates that the computer system is at the onset of degradation; andif so, performing a remedial action. 11. The computer-readable storage medium of claim 10, prior to receiving the variance for the time series for the monitored telemetry variable, the method further comprises:receiving the time series for the monitored telemetry variable; andcalculating the variance of the time series for the monitored telemetry variable. 12. The computer-readable storage medium of claim 10, wherein calculating the first-difference function of the time series involves, for each time point within the time series, subtracting a value of the time series at a previous time point from the value of the time series at a present time point. 13. The computer-readable storage medium of claim 12, wherein the method further comprises dividing the result of the subtraction by the value of a length of a time interval between the previous time point and the present time point. 14. The computer-readable storage medium of claim 10, wherein calculating the residual function for a time series involves:for each time interval in the time series,calculating a running average of values for the time series up to and including a present time interval; andsubtracting the running average from a value of the time series at the present time interval. 15. The computer-readable storage medium of claim 10, wherein determining whether the first-difference function indicates that the computer system is at the onset of degradation involves determining whether the first-difference function exceeds a specified threshold. 16. The computer-readable storage medium of claim 10, wherein determining whether the first-difference function indicates that the computer system is at the onset of degradation involves:performing a Sequential Probability Ratio Test (SPRT) on the first-difference function; anddetermining whether the SPRT generates an alarm. 17. The computer-readable storage medium of claim 16, wherein the SPRT can include one or more of:a positive variance first-difference test, which generates an alarm if the first-difference function for the variance of the time series for the monitored telemetry variable is increasing; anda negative variance first-difference test, which generates an alarm if the first-difference function for the variance of the time series for the monitored telemetry variable is decreasing. 18. The computer-readable storage medium of claim 10, wherein performing the remedial action can involve performing one or more of:recording a time when the onset of degradation occurred;notifying a system administrator that the computer system is at the onset of degradation;shutting down the computer system;backing up data stored on the computer system;failing-over to a redundant computer system; andreplacing one or more components which are at the onset of degradation. 19. An apparatus that monitors the health of a computer system, comprising:a receiving mechanism configured to receive a variance of a time series for a monitored telemetry variable within the computer system,a residual-calculation mechanism configured to calculate a residual function of the variance of the time series for the monitored telemetry variable;a difference-calculation mechanism configured to calculate a first-difference function of the variance of the time series from the residual function;a degradation-detection mechanism configured to determine whether the first-difference function indicates that the computer system is at the onset of degradation; anda remedial-action mechanism, wherein if the degradation-detection mechanism determines that the computer system is at the onset of degradation, the remedial-action mechanism is configured to perform a remedial action.
	description	This application claims the priority, under 35 U.S.C. §119, of German Patent Application DE 10 2010 035 955.6, filed Aug. 31, 2010; the prior application is herewith incorporated by reference in its entirety. The invention relates to a fuel element storage and cooling configuration including a fuel element storage pool and a cooling system. It is generally known that preferably uranium-containing fuel elements are used as nuclear fuel in nuclear power generation plants. Those fuel elements have a mostly rod-like shape, depending on the type of reactor, with a typical pressurized-water reactor of 1200 MW output requiring approximately 190 such fuel elements for one core load. Each year, approximately 20% of all of the fuel elements must be replaced by new ones after their respective burnup, depending on the usage type of the nuclear reactor. For that purpose, the respective fuel elements are removed from the reactor core, which is disposed inside a reactor containment, by way of a corresponding robot-type apparatus and transferred to a neighboring fuel element storage pool, also known as a spent fuel pool, which is likewise located in the reactor containment, and temporarily stored there. Even if the fuel elements are burnt up in such a way that they can no longer be used in a reactor, they still initially have a not inconsiderable decay power that is associated with a corresponding heat generation, which certainly falls within the MW range. In order to protect the operating staff and the environment against radioactivity that occurs during that process, and at the same time to ensure heat dissipation of the temporarily stored fuel elements, the fuel element storage pool, or the spent fuel pool, is filled with pure water in such a way that the actual fuel element storage location at the bottom of the spent fuel pool is surrounded to the sides and toward the top by several meters of water. For the purpose of maintaining the reactor, the entire core load of fuel elements is temporarily stored in the spent fuel pool, if so required, in such a way that access to the reactor core is ensured. Depending on the storage levels, thermal outputs of 10 MW and more occur in the spent fuel pool, which could bring the water therein to the boil within a matter of a few hours. However, that must be prevented for reasons of safety under all circumstances. Therefore, spent fuel pools have cooling systems which guide the generated thermal output, using heat exchangers, to a heat sink, that is to say to a region or a device that can absorb heat. According to the prior art, cooling systems for spent fuel pools in the reactor sector are realized without exception by way of using active components, especially pumps. The use of active components allows the efficiency of a heat exchanger, which is likewise a conventional measure for this purpose, to be increased in an advantageous manner, but the disadvantage is that if the active components fail, the spent fuel pool can possibly overheat over a time period in the hour range. Even if correspondingly frequent maintenance and the provision of corresponding emergency power systems nearly rule out the likelihood of such an event occurring, it is desirable, due to the extremely high safety demands in the reactor sector, to further increase the reliability of cooling systems in spent fuel pools, particularly under extreme conditions. It is accordingly an object of the invention to provide a fuel element storage and cooling configuration, which overcomes the hereinafore-mentioned disadvantages of the heretofore-known devices of this general type and which makes an improved cooling system available for spent fuel pools. With the foregoing and other objects in view there is provided, in accordance with the invention, a fuel element storage and cooling configuration, comprising a fuel element storage pool, a heat sink, a cooling system including at least one first heat exchanger disposed in the fuel element storage pool and having a highest point and at least one second heat exchanger disposed in the heat sink at a distance from the at least one first heat exchanger above the highest point of the at least one first heat exchanger, and a pipe system at least partially filled with a flowable coolant and interconnecting the at least one first and the at least one second heat exchangers to form a closed circuit. The pipe system ensures natural circulation of the coolant and thus heat transport from the fuel element storage pool to the heat sink, without a pump apparatus, upon a temperature increase of the at least one first heat exchanger relative to the at least one second heat exchanger. The core concept of the invention is to completely dispense with active components in the cooling system and to thus further reduce the likelihood of their failure. A typical active component is a pump which, for example, pumps a coolant in an intermediate circuit between two heat exchangers. Such intermediate circuits are advantageous in nuclear facilities for reasons of safety in order to prevent contaminated water from leaking out of the fuel element storage pool into the environment under all circumstances. Specifically, the height difference between the first and the second heat exchanger makes it possible to achieve so-called natural circulation of the flowable coolant. The coolant, which is preferably introduced into the first heat exchanger from below, has a lower temperature than the heat exchanger and heats up as it flows through the internal heat-exchanger pipes thereof, that is to say it takes on at most the temperature of the water surrounding the first heat exchanger. In the process, the coolant expands in such a way that it becomes less dense, that is to say it is lighter per unit volume than the following, colder coolant. The outlet for the outflow of the coolant from the first heat exchanger is provided in the upper region thereof in such a way that the coolant with the higher specific density which follows pushes the heated coolant with lower specific density upward. Disposed above the first heat exchanger is the second heat exchanger which cools the coolant in such a way that it flows downward again in the circuit, with correspondingly higher specific density, to the first heat exchanger. In order to achieve a desired cooling output, the diameter of the pipes and possibly the contact area of the heat exchangers can be increased, if appropriate. This possibly increased outlay in terms of construction is, however, justified due to an increased measure of safety. Thus, a cooling system is provided for dissipating heat from the spent fuel pool, which advantageously dispenses with any active components. However, the invention is not at all restricted to a purely liquid coolant. In accordance with another feature of the invention, the at least one first heat exchanger is configured as an evaporator and the at least one second heat exchanger is configured as a condenser. The coolant should be selected in terms of its boiling and condensation points under the given pressure and temperature conditions, in such a way that, if it is introduced as a liquid into the first heat exchanger, or evaporator, it evaporates there and has a considerably higher cooling effect due to the evaporation energy required. The now gaseous coolant then flows likewise in the direction of the second heat exchanger, or condenser, which is situated at a higher location. In this case, the coolant reverts to the liquid state while giving off thermal energy and can be guided through the circuit under the force of gravity, into the evaporator, without the use of a pump. The cooling output of the fuel element storage pool and cooling system according to the invention is thus increased in an advantageous manner. In accordance with a further feature of the invention, the flowable coolant is a refrigerant. Like coolants, refrigerants transport enthalpy. The difference is that a refrigerant can do so even against a temperature gradient in such a way that the ambient temperature can even be higher than the temperature of the object to be cooled. In contrast thereto, a coolant can only transport the enthalpy along the temperature gradient to a place of lower temperature. According to the invention, refrigerants that have proven suitable for this purpose are, for example, the group of refrigerants having two carbon atoms, which is known under the designation R1xx, with xx being a placeholder for the numbering of the various variants, as is known to a person skilled in the art, however. The refrigerant R134a with the chemical formula C2H2F4 has proven particularly suitable. In this way, the cooling output is increased once again. In accordance with an added feature of the invention, the at least one first heat exchanger is disposed in such a way that it is suspended, for example through the use of a suitable holding apparatus at the pool edge. In this way, the flow can pass more easily through the now vertical cooling pipes in an advantageous manner. It must be ensured that the first heat exchanger during operation of the spent fuel pool is located completely below the surface of the now necessary water filling in order to ensure maximum heat exchange. The two necessary feed pipelines can likewise be guided to the heat exchanger from above over the pool edge, in such a way that there is no need for a through-hole in the pool wall. In this way, very simple installation of a cooling system, which can if appropriate also be retrofitted without a great amount of outlay, is ensured. In accordance with an additional feature of the invention, it proves advantageous to place a baffle on an outer face of the at least one first heat exchanger. This baffle should be configured in such a way that the water heated by the heat generated by the stored fuel elements, which water rises up from the fuel element storage site for temperature reasons, is guided in the direction of the first heat exchanger in order to enable an improved heat exchange. In accordance with yet another feature of the invention, in a particularly preferred configuration, the second heat exchanger is disposed 3 m-5 m above the highest point of the first heat exchanger. Such a height difference proves particularly suitable for building up so much pressure of the condensed or cooled coolant, under the force of gravity, that it flows from the second heat exchanger back into the first heat exchanger even over a relatively long horizontal distance, for example a few tens of meters or even 100 m. Such a distance results, for example, from the distance from the spent fuel pool to the cooling tower. The pipelines between the first and the second heat exchanger, and in particular the return line for the cooled or condensed coolant, should preferably be configured to have a continuous slope. This ensures a barrier-free natural circulation of the coolant. In accordance with yet a further feature of the invention, closure devices, which are activated in the event of a sudden pressure drop and stop circulation of the coolant, are provided in a point-by-point manner inside the closed circuit. This is intended to ensure that, in the event of a leak in the cooling circuit, no contaminated water passes from the spent fuel pool through the heat exchanger into the cooling circuit or from there into the environment. In accordance with yet an added feature of the invention, in a particularly preferred invention variant, a natural draft dry cooling tower is used as the heat sink. This is because a natural draft dry tower brings about forced cooling of the second heat exchanger likewise without active components, which should be avoided according to the invention. In accordance with yet an additional feature of the invention, a plurality of cooling systems, which are independent from one another at least with respect to the circuit, is provided. The use of a plurality of similar cooling systems, which summarily make available a desired cooling output, reduces the consequences of the failure of a system in an advantageous manner. In order to ensure the independence of the systems, at least the cooling circuits must be configured independently of one another in order to continue to be operational at least with reduced cooling output in the event of a leakage in a cooling circuit. Placing a plurality of second heat exchangers or condensers in the same cooling tower, however, is acceptable since the likelihood of a cooling tower failing, even measured against the most stringent safety requirements in the nuclear engineering sector, is particularly low. In accordance with again another feature of the invention, in order to increase the operational reliability further, it is expedient to configure the cooling systems in an overredundant manner, that is to say to summarily make available more than the maximum needed cooling output, wherein the output reserve includes at least the individual output of the most efficient individual cooling system. More preferably, the cooling systems should be of identical configuration, wherein overall the failure of two individual cooling systems can be compensated for. Another possibility for increasing the operational reliability is to configure the cooling systems in a diverse manner. In this case, an output excess with respect to the intended maximum cooling output is likewise made available, but the various cooling systems operate according to different cooling principles. It is thus possible to provide, for example, some of the required cooling output by way of active cooling systems, which the invention actually tries to avoid, where another part of the cooling output is made available by way of passive cooling systems according to the invention. Due to the different nature of the types of cooling, increased reliability is ensured, at least for some of the maximum cooling output. In accordance with a concomitant feature of the invention, the cooling system is configured for a cooling output of 5 MW-30 MW. This corresponds to the cooling output to be intended for the spent fuel pool of a typical reactor in the output range of 600 MW to 1500 MW. Other features which are considered as characteristic for the invention are set forth in the appended claims, noting that further advantageous possible embodiments can be gathered from the further dependent claims. Although the invention is illustrated and described herein as embodied in a fuel element storage pool with a cooling system, it is nevertheless not intended to be limited to the details shown, since various modifications and structural changes may be made therein without departing from the spirit of the invention and within the scope and range of equivalents of the claims. The construction and method of operation of the invention, however, together with additional objects and advantages thereof will be best understood from the following description of specific embodiments when read in connection with the accompanying drawings. Referring now to the figures of the drawings in detail and first, particularly, to FIG. 1 thereof, there is seen an example of a configuration 10 having a fuel element storage pool 12 with a cooling system. A fuel element storage location 38 is centrally disposed at a pool bottom of the fuel element storage pool 12 which has, for example, a diameter of 10 m. The storage location 38 has a multiplicity of stack-like storage positions for rod-like fuel elements, for example several hundred thereof. The fuel element storage pool 12 is filled with water 42, having a fill level 40 which is just below a pool edge 44. Two first heat exchangers 14, 16, which are suspended and force-lockingly connected to the pool edge through the use of respective holding elements 46, are likewise disposed in the fuel element storage pool 12 below the fill level 40. A force-locking connection is one which connects two elements together by force external to the elements, as opposed to a form-locking connection which is provided by the shapes of the elements themselves. The highest point of the first heat exchangers 14, 16 is indicated by a reference line 24. This reference line 24 is a height reference point for second heat exchangers 18, 20 which are to be disposed at a higher location. In a real or actual configuration, however, a plurality of first heat exchangers must be provided, in particular for redundancy reasons. The fuel elements (not shown in the figure) which are stored in the fuel element storage location 38 output a thermal output, for example 16 MW, into the water 42, which is indicated by an arrow 34 and is ultimately transferred into the first heat exchangers 14, 16 in the form of a thermal input. In this regard, two guiding plates or baffles 74, which may be disposed on or near an outer surface of the first heat exchangers 14, 16, are provided. Two pipeline parts 30, 32, which are necessary for forming a first cooling circuit, are indicated for the first heat exchanger 14. However, a cooling circuit must, of course, also be provided for each further heat exchanger in a preferably independent form, as is indicated by partially illustrated sections of pipelines 78 of a second cooling circuit. Thus, a first independent cooling system 14, 18, 30, 32 as well as a second independent cooling system 16, 20, 78, are provided. It is understood that several cooling systems may be provided in an over-redundant manner. The pipeline parts 30, 32 are guided over the pool edge and connected to the first heat exchanger 14, so that a through-hole in the pool wall is advantageously avoided. The pipeline parts 30, 32 are guided, preferably with a continuous slope in the direction of the spent fuel pool 12, from the spent fuel pool 12 through a reactor containment wall 48, which surrounds a non-illustrated reactor and the spent fuel pool 12, to the outside into the center of a natural draft cooling tower 22 serving as a heat sink. This natural draft cooling tower 22 is disposed, for example, at a distance of 50 m from the spent fuel pool 12 (although this is not shown to scale). The two second heat exchangers 18, 20, which are preferably configured as condensers, are disposed in the cooling tower 22, and the height of the lower edge of the heat exchangers 18, 20 is indicated by reference numeral 26. There is a height difference, for example of 3 m-5 m, which is indicated by an arrow 28, between the upper edge 24 of the first heat exchangers 14, 16, which are preferably configured as evaporators, and the lower edge 26 of the heat exchangers 18, 20. Each condenser 18, 20 is to be configured as an independent condenser field with separate feed lines 30, 32, in order to form a plurality of mutually independent cooling circuits. These cooling circuits are filled with a coolant. Due to the thermal input 34 into the evaporator 14, the coolant evaporates and is guided in the gaseous state through the second part 32 of the pipeline system to the first condenser 18, where it condenses. A thermal output 36, which is output in the process, is transferred to a heat sink, in this case the cooling tower 22. The condensed coolant is returned in a continuous slope, under the force of gravity, through the first part 30 of the pipeline system, to the evaporator 14, and the cooling circuit is thus closed. FIG. 2 shows an exemplary circuit with a coolant 50. An evaporator 52 and a condenser 54 are connected through a first part 60 and a second part 62 of a pipeline system to form a closed cooling circuit. The evaporator 52 and the condenser 54 have a multiplicity of individual cooling pipes, which at their two ends are brought together in each case through a collecting connection to form a respective pipe connection. The multiplicity of individual cooling pipes serves to increase the contact area, thus resulting in an increased cooling output. The evaporator 52 is filled with a coolant 56 up to a height which is indicated by a dashed line. Due to a thermal input 70 into the evaporator 52, the coolant 56 evaporates with absorption of a thermal output and is guided in gaseous form through the pipeline part 62 to the condenser 54. In this case, the coolant condenses into a liquid state 58 while giving off a thermal output 72 and is returned to the evaporator through the pipeline part 60 under the force of gravity. The condenser 54 is disposed with its lowermost point 66 at a vertical distance 68 above an uppermost point 64 of the evaporator 52, in such a way that a natural circulation of the coolant 56, 58 is ensured, without the need for active components such as a pump. Valves or closure devices 76 are provided in a point-by-point manner inside the closed circuit 50. The closure devices 76 are activated in the event of a sudden pressure drop to stop circulation of the coolant 56, 58. Whereas FIG. 2 shows only a passive cooling system, FIG. 3 shows an active cooling system which has a pump 82 for coolant 80 and can be provided in addition to the passive cooling system, thus providing diverse cooling systems.
	summary
041359739	claims	1. A nuclear reactor system including a vessel, fuel assemblies positioned therein, an inlet and an outlet for circulating a coolant in heat transfer relationship with said fuel assemblies, and a closure head disposed on said vessel in a fluid-tight relationship, said closure head comprising: a substantially cylindrical outer riser assembly disposed on said reactor vessel and having a forked member disposed therein; a substantially cylindrical inner riser assembly disposed in said outer riser assembly and defining an annulus between said riser assemblies; an elastomer ring having a corrugated contact surface disposed in said forked member and extending near said annulus for sealing said annulus; a tubular inflatable seal disposed in said forked member near said elastomer ring for expanding against said elastomer ring when inflated thereby causing said elastomer ring to radially expand against said inner riser assembly thus sealing said annulus; a metallic stem attached to said tubular inflatable seal and a passageway connected between said metallic stem and a gas source for conducting a gas from said gas source to said tubular inflatable seal for inflating said tubular inflatable seal; a check valve disposed in said passageway for selectively preventing deflation of said inflatable seal; and a conduit disposed in said forked member and having one end disposed in fluid communication with the side of said tubular inflatable seal that is opposite said elastomer ring and having the other end disposed in fluid communication with said annulus below said tubular inflatable seal whereby increased pressure in said annulus below said tubular inflatable seal will be transmitted through said conduit and against said tubular inflatable seal, thus further expanding said tubular inflatable seal and said elastomer ring and thus enhancing contact between said corrugated contact surface and said inner riser assembly.
051130785	claims	1. A radiation shielding structure comprising: a radiation shielding panel including a lead transparent plate and a nonlead transparent plate which is laminated on at least one side of said lead transparent plate; and a gasket which is fitted to the outer peripheral edge of said radiation shielding panel to hermetically seal the area between said radiation shielding panel and a panel mounting portion, wherein the outer peripheral edge of said radiation shielding panel is formed with a gasket fitting recess which is fitted with a positioning projection that is integrally formed on the inner peripheral edge of said gasket. a radiation shielding panel including a lead transparent plate and a nonlead transparent plate which is laminated on at least one side of said lead transparent plate; and a gasket which is fitted to the outer peripheral edge of said radiation shielding panel to hermetically seal the area between said radiation shielding panel and a panel mounting portion, wherein the outer peripheral edge of said gasket is integrally formed with contact projections for improving the adhesion between said gasket and said panel mounting portion. 2. A radiation shielding structure comprising: 3. A radiation shielding structure according to claim 2, wherein the outer peripheral edge of said radiation shielding panel is formed with a gasket fitting recess which is fitted with a positioning projection that is integrally formed on the inner peripheral edge of said gasket.
	claims	1. A system comprising:a radiation containment chamber;an isolator connected to the radiation containment chamber;a rotating transfer door positioned between the radiation containment chamber and the isolator and including a cavity for receiving a radionuclide generator column assembly,the transfer door rotatable between a first position, in which the cavity is open to the radiation containment chamber, and a second position, in which the cavity is open to the isolator; andan antimicrobial vapor generator connected to the isolator, wherein the transfer door is adapted to rotate while antimicrobial vapor is introduced into the isolator by the antimicrobial vapor generator. 2. The system of claim 1 further including an enclosure constructed of radiation shielding material, the transfer door located within an enclosure cavity defined by the enclosure, the enclosure defining a first access opening that provides access to the enclosure cavity from the radiation containment chamber, and a second access opening that provides access to the enclosure cavity from the isolator. 3. The system of claim 2, wherein the enclosure includes an air return in fluid communication with the enclosure cavity and connected to a recirculation device that generates a localized negative pressure within the enclosure cavity to draw antimicrobial vapor into the enclosure cavity. 4. The system of claim 2 further including a sealing door positioned between the radiation containment chamber and the transfer door, the sealing door movable between an open position and a closed position in which the sealing door seals the first access opening to isolate the isolator and the transfer door from the radiation containment chamber. 5. The system of claim 2, wherein the enclosure cavity permits air flow from the isolator to the radiation containment chamber, and wherein the transfer door restricts air flow between the isolator and the radiation containment chamber to a flow path defined between the transfer door and the enclosure. 6. The system of claim 2 further including:a bearing assembly rotatably supporting the transfer door, the bearing assembly including a lower bearing disposed between a base of the transfer door and a bottom of the enclosure, and an upper bearing disposed between a top of the transfer door and an upper support shaft connected to a top of the enclosure. 7. The system of claim 1 further including a motor connected to the transfer door, the motor configured to continuously rotate the transfer door while antimicrobial vapor is introduced into the isolator by the antimicrobial vapor generator. 8. The system of claim 1, wherein the transfer door includes a base and a radiation shield mounted to and extending upwards from the base, the radiation shield defining the cavity for receiving a radionuclide generator column assembly. 9. The system of claim 8, wherein the radiation shield is constructed of at least one of lead, tungsten, and depleted uranium. 10. The system of claim 8, wherein the transfer door has a weight of at least 400 pounds and is mounted on a lower bearing adapted to support the majority of the transfer door weight. 11. The system of claim 8, wherein the radiation shield has a thickness profile such that the radiation shield provides no less than 6 inches of radiation shielding between the radiation containment chamber and the isolator, irrespective of a rotational position of the transfer door. 12. The system of claim 1, wherein the isolator is positively pressurized relative to the radiation containment chamber. 13. The system of claim 1, wherein the radiation containment chamber encloses a first clean room environment, and the isolator encloses a second clean room environment having a higher clean room classification than the first clean room environment. 14. A method of sanitizing an isolator connected to a radiation containment chamber of a system for producing radionuclide generators, the system including a rotating transfer door for transferring radionuclide generator column assemblies between the isolator and the radiation containment chamber, the rotating transfer door positioned between the radiation containment chamber and the isolator and including a cavity for receiving the radionuclide generator column assembly, the method comprising:introducing an antimicrobial vapor into the isolator with an antimicrobial generator connected thereto; androtating the transfer door while the antimicrobial vapor is circulated within the isolator, wherein rotating the transfer door includes rotating the transfer door between a first position, in which the cavity is open to the radiation containment chamber, and a second position, in which the cavity is open to the isolator. 15. The method of claim 14, wherein the transfer door is positioned within an enclosure cavity defined by an enclosure, the method further including generating a localized negative pressure within the enclosure cavity to draw the antimicrobial vapor within and around the transfer door. 16. The method of claim 15, wherein the enclosure includes an air return fluidly connected to the enclosure cavity, the method further including circulating the antimicrobial vapor into the enclosure cavity and through the air return. 17. The method of claim 14, further comprising sealing the isolator and the transfer door from the radiation containment chamber prior to introducing the antimicrobial vapor into the isolator. 18. The method of claim 17, wherein sealing the isolator and the transfer door from the radiation containment chamber includes moving a sealing door from an opened position to a closed in which the sealing door isolates the transfer door and the isolator from the radiation containment chamber. 19. The method of claim 14, wherein introducing an antimicrobial vapor into the isolator includes introducing vaporized hydrogen peroxide into the isolator. 20. The method of claim 14, wherein rotating the transfer door includes continuously rotating the transfer door while the antimicrobial vapor is circulated within the isolator. 21. The method of claim 14, wherein rotating the transfer door includes rotating the transfer door at least one complete revolution while the antimicrobial vapor is circulated within the isolator.
051397323	claims	1. Process for extracting a heating rod (3) having deformations from a pressurizer casing (2) of a pressurized-water nuclear reactor, which has an axis of symmetry and in which the heating rods (3) are held in an axial arrangement by spacer plates (4a, 4b) and pass through a bottom (2a) of the casing (2) inside sleeves (12), wherein the heating rod (3c, 3d) is cut inside the pressurizer casing in at least one zone by means of a cutting operation controlled remotely, and in that at least one portion of the rod (3c, 3d) is extracted by way of an inspection port (7) of the casing (2). 2. Extraction process according to claim 1, wherein the heating rod (3c) is cut in the vicinity of at least one spacer plate (4a, 4b). 3. Extraction process according to claim 1, wherein the heating rod is cut in the vicinity of the end of a passage sleeve (12) located inside the casing (2) of the pressurizer (1). 4. Extraction process according to claim 1, wherein at least one portion of the heating rod (3c, 3d) is extracted from outside the casing (2) of the pressurizer (1) by pulling on its end located outside a corresponding passage sleeve (12). 5. Device for extracting a heating rod (3) having deformations from a casing (2) of a pressurizer (1) of a pressurized-water nuclear reactor, which has an axis of symmetry and in which the heating rods (3) are held in an axial arrangement by spacer plates (4a, 4b) and pass through the bottom of the casing (2) inside sleeves (12), said device comprising at least one remotely controlled cutting mechanism (25, 25', 35) comprising a cutting means (30) and means (31a, 31b, 33a, 33b) for moving the device between two concentric rows of heating rods (3). 6. Device according to claim 5, wherein the device (25, 25') comprises a central element (28) and at least one end slide (31a, 31b) movable in translational motion relative to the central element (28) and associated with driving means for movement in translational motion, and remotely controllable retractable grasping fingers carried by the central element (28) and each of the slides (31a, 31b) and capable of engaging into and of being clamped in water passage orifices (27) passing through the spacer plates (4a, 4b) and arranged equidistantly between concentric rows of heating rods (3) held in position by the spacer plates (4a, 4b). 7. Device according to claim 5, comprising a central element and at least one end slide mounted movably in translational motion relative to the central element and associated with remotely controlled means of movement in translational motion, and remotely controllable retractable grasping forks (37, 37', 38a, 38b) carried by the central element and the slides and capable of coming into engagement with the successive heating rods (3) of a circular row of heating rods immediately above the passage sleeves (12) of the heating rods (3). 8. Device according to claim 5, further comprising, furthermore, means for lifting and placing at least one remote-controlled cutting mechanism (25, 25', 35) at the end of a space located between two concentric rows of heating rods (3). 9. Device according to claim 8, wherein the means for handling the remote-controlled cutting mechanism (25, 25', 35) comprise a structure (41) fastened in an axial arrangement inside the pressurizer casing (20, a transfer rail (42) mounted rotatably about an axially directed axis in a transverse arrangement relative to the structure (41), at least one ramp (43, 44, 45) for supporting and guiding at least one remote-controlled cutting mechanism (25, 25', 35) and mounted pivotably on the transfer rail (42), and means (46, 46',47) for placing at least one cutting mechanism (25, 25', 35) on a support and guide ramp (43, 44, 45).
	description	The present invention relates to the nuclear sector, and more specifically to the area of BWR (Boiling Water Reactor) technology vessels. These plugs are used in the decontamination tasks of recirculation loops (channeled plugs) and in maintenance tasks on the valves of the recirculation loops (blind plugs). The blind plugs of jet pumps are used to seal the five outlets of the nozzle of the jet pump such that the recirculation loop can be isolated from the reactor during the maintenance tasks on the discharge valves of the manual or automatic recirculation pump. The channeled plugs are used to close the decontamination circuit of the recirculation loops, preventing the cleaning solution from being dispersed in the reactor. In other words, they establish the closed circuit. There are currently plugs marketed by other companies to perform both functions, channeling and plugging, but with different drawbacks that are resolved by these plugs. Firstly, the equivalent pitch diameter of plugs marketed up until now is ¼″ (13.7 mm.) whereas those herein developed have a pitch diameter of 1″ (33 mm.) The marketed plugs are mounted in several sub-assemblies, whereas the device of the invention is a single part, i.e., in a single assembly. This device includes two covers carrying plugs, such that one of them serves to plug the even nozzles, whereas the other one closes the odd nozzles of each jet pump. The combination of both plug units (blind and channeled) allows adjusting the flow rate values in decontamination, such that the flow and head losses are forced by determined jet pumps and prevented by others. Each of these covers is mounted on a common base and at the end of respective arms articulated in the central area, which are actuated by mechanical or pneumatic means. The installation is carried out by applying the corresponding plug on the nozzles of the jet pump, once the base of the plug is supported on the mixer part of the jet pump through a guide which allows centering the plug device therein. The device is also supported in the jet pump in respective reactions, a lower reaction serving as a support both when introducing the plug in the pump and when demounting same, and another upper reaction serving to allow the closure of the plug against the nozzles of the jet pump, which are located on the plug. Each of the covers carrying plugs is mounted in the corresponding arm in a pivoting manner on a central point, which allows the self-alignment of the plane of the plugs with the plane of the nozzles to be blocked. The placement of an adjusting screw between the arm supporting each cover and the support guide of the base at the opening of the mixer of the jet pump has also been provided, which adjusting screw allows assuring the parallelism of the sealing surface of the plug with the plane of the nozzles to be plugged of the jet pump. The covers carrying plugs are optionally attached or are a single cover, incorporating in this case the five plugs in a suitable arrangement and are actuated by means of a single arm and clamping screw. Furthermore, two models referred to as a blind plug and a channeled plug have been developed in the inventive unit of this device. The use of each of them is described below: The blind plug only serves to seal the five nozzles of the jet pump. The channeled plug enables fluid circulation during the decontamination tasks with a section equivalent to 32 mm. To do so it seals the three odd nozzles of the jet pump, while at the same time it conducts the two remaining nozzles to a flow outlet. As can be seen in the mentioned figures, this device includes five plugs (6) mounted in a suitable arrangement to close the five outlets of the nozzle of the jet pump in order to allow isolating the recirculation loop from the reactor during the maintenance tasks in the discharge valves of the recirculation pump. As can be seen in the figures, the plugs (6) are mounted with the following arrangement: two plugs on cover (1) and three on cover (2), in all cases the three plugs located on cover (2) are blind, whereas the two plugs located on cover (1) in the example shown in FIGS. 1 and 2 are blind, but in the example shown in FIGS. 3 and 4 they are channeled, as will be explained below. Each cover (1-2) is mounted at the end of an arm (4) articulated in a common base (3) and actuated from a screw (5), such that starting from a position in which both covers are dropped downwards, the latter are introduced in the jet pump under the five outlets of the nozzle and are actuated until being placed in the horizontal position depicted in FIGS. 1 and 3, in which all the outlets would be blocked. The base (3) of the device is supported on the mixer part of the jet pump (10), the guide (12) being the part that centers the plug device in two phases, a first approximation phase and another more precise phase performed with the adjusting screw (15) which serves to assure the parallelism of the sealing surface of the plug with the plane of the nozzles to be plugged of the jet pump. The pivoting mounting of the covers (1-2) in the arms (4) at an intermediate point (7) allows the self-alignment of the plane of the plugs (6) with the plane of the nozzles to be blocked. The lower reaction (8) serves as support both when introducing the plug in the pump and when demounting same. The upper reaction (9) serves to allow the closure of the plug against the nozzles of the jet pump which are located on the plug. The device can also be mounted on a single arm cover and support whereby it would be more versatile. FIGS. 3 and 4 show the operation of the channeled plug. The flow inlet (11) is through two of the five nozzles of the jet pump. The outlet (14) is through the plug (13), which is channeled in this case. Having sufficiently described the nature of the invention as well as a preferred embodiment, it is hereby stated for all relevant purposes that the materials, shape, size and arrangement of the described elements can be modified providing that this does not involve an alteration of the essential features of the invention which are claimed below.
	description	The present invention is directed to improved image reconstruction methods for cone beam ROI imaging of long objects with an area detector using a series of circular scans, wherein essentially the entire detector area is utilized for all source positions. More specifically, preferred embodiments of the invention comprise extensions of the data acquisition and image reconstruction techniques disclosed in the above-incorporated U.S. Pat. No. 5,383,119 to improve utilization of the detector. Rather than filter out (via a mask) the portion of the cone beam projection data (which is captured by the circular scans on the ROI boundaries) that lies outside the ROI, the cone beam data outside the ROI is preferably processed using the methods disclosed in the reference by Kudo, et al., entitled xe2x80x9cAn Extended Completeness Condition for Exact Cone-Beam Reconstruction and its Applicationxe2x80x9d, IEEE Nucl. Sci Symp. Conf. Record, 1994, Norfolk, Va., pp 1710-1714 (referred to herein as xe2x80x9cKudoxe2x80x9d), which is incorporated herein by reference. Further, the cone beam data within the ROI is acquired and processed using the methods disclosed in the above-incorporated U.S. Pat. No. 5,383,119. Image reconstruction methods according to the invention provide reconstruction of an extended ROI, i.e., a ROI that includes a portion of the object that lies beyond the boundary extents of the circular scans. Advantageously, image reconstruction methods according to the invention enable virtually full utilization of the detector, while providing reconstruction of a longer (extended) ROI with the same projection data acquired by the scanning trajectories for the method of U.S. Pat. No. 5,383,119, within the same scan time. In other words, by optimally utilizing the detector, a longer volume can be reconstructed for the same data acquisition scan time. Details of image reconstruction methods according to the invention will now be discussed with reference to FIGS. 5-9. FIG. 5 generally illustrates a cone beam data acquisition process for an image reconstruction method according to an embodiment of the present invention. More specifically, FIG. 5(a) illustrates a given source position on the upper circular scan 21, wherein an xe2x80x9cextendedxe2x80x9d portion of the long object 20 (which is located above the desired ROI xcexa9o in the figure) is reconstructed using cone beam projection data captured on the top half (T) of the detector D. FIG. 5(b) illustrates the utilization of the detector D for the upper circular scan depicted in FIG. 5(a). As shown, the cone beam projection data on the upper half (T) of the detector D is not masked (filtered) out but rather fully utilized. The cone beam projection data that is captured on the detector portion T (i.e., v greater than 0) is processed using the methods described in the Kudo reference, whereas the cone beam projection data that is captured on the detector portion (B) (i.e., for vxe2x89xa60)is processed for data combination with the bottom circular scan path data using the methods disclosed in the above-incorporated U.S. Pat. No. 5,383,119. This is to be contrasted with the method depicted in FIGS. 3 and 4b, for example, wherein the cone beam projection data for v greater than 0 is filtered out via mask M such that the cone beam projection data on the top half (T) of detector D is not utilized for image reconstruction. A similar cone beam data acquisition process is applied for the lower circular scan 22 as depicted in FIG. 6. In particular, FIG. 6(a) illustrates a given source position on the lower circular scan 22, wherein an xe2x80x9cextendedxe2x80x9d portion of the long object 20 (which is located below the desired ROI xcexa9o in the figure) is reconstructed using cone beam projection data captured on the bottom half (B) of the detector D. FIG. 6(b) illustrates the utilization of the detector D for the lower circular scan 22 depicted in FIG. 6(a). As shown, the cone beam projection data on the lower half (B) of the detector D is not masked (filtered) out, but rather fully utilized. The cone beam projection data that is captured on the bottom half (B) of the detector (i.e., v less than 0) is processed using the methods described in the Kudo reference, whereas the cone beam projection data that is captured on the upper half (T) of the detector D (i.e., for vxe2x89xa70)is processed for data combination with the upper circular scan path data using the methods disclosed in the above-incorporated U.S. Pat. No. 5,383,119. This is to be contrasted with the method depicted in FIGS. 3 and 4c, for example, wherein the cone beam projection data for v less than 0 is filtered out via mask M such that the cone beam projection data on the bottom half (B) of detector D is not utilized for image reconstruction. FIG. 7 illustrates an extended ROI, comprising a ROI xcexa9o, and extended portions xcexa9u and xcexa9d, which is reconstructed using cone beam data that is acquired from two circular scan paths 21 and 22 (FIGS. 5 and 6), and a connecting line scan L, using an image reconstruction method according to an embodiment of the invention. When the entire top circular scan data is processed using the method depicted in FIG. 5, the portion of the cone beam image with v greater than 0 yields reconstruction of the volume xcexa9u shown in FIG. 7. Similarly, when the entire bottom circular scan data is processed using the method depicted in FIG. 6, the portion of the cone beam image with v less than 0 yields reconstruction of the volume xcexa9d shown in FIG. 7. The total reconstructed volume with the two circular scan paths is extended from xcexa9o to xcexa9u ∪xcexa9o ∪xcexa9d. Further details of the data acquisition and processing method for reconstructing the extended ROI depicted in FIG. 7, will now be explained in further detail. As discussed above, the method disclosed in the above-incorporated U.S. Pat. No. 5,383,119 is extended to perform a data acquisition and image reconstruction process that improves utilization of the detector and provides reconstruction of an extended ROI, i.e., a reconstructed ROI which is larger that the reconstructed ROI obtained using the method U.S. Pat. No. 5,383,119, but with the same scanning paths and scanning time as the conventional method. To improve the utilization of the detector, the cone beam data in the v greater than 0 portion (see FIG. 5(b)), that is obtained from the upper circular scan, together with the cone beam data from a line scan, are used to reconstruct the portion xcexa9u of the long object that extends from the end of xcexa9o using the Kudo method. Similarly, using the Kudo method, the cone beam data in the v less than 0 portion (see FIG. 6(b)), that is obtained from the lower circular scan, together with the cone beam data from the line scan, are used to reconstruct the portion xcexa9d of the long object that extends from the end of xcexa9o. Details of a preferred method for processing cone beam data for the extended portions, xcexa9u and xcexa9d, using the Kudo method will now be provided. First, the portion of the cone beam image data that corresponds to the extended portions, i.e., the cone beam data for v greater than 0 for the upper circular scan, and the cone beam data for v less than 0 for the lower circular scan, is processed using the well-known xe2x80x9cFeldkampxe2x80x9d algorithm, which is disclosed for example in the article by Feldkamp, et al., entitled xe2x80x9cPractical Cone-Beam Algorithmxe2x80x9d, Journal of the Optical Society of America, Vol. 1, pp. 612-619, 1984, which is incorporated herein by reference. Essentially, with the Feldkamp method, a ramp filtering process is applied to the relevant projection data of the circular scans. Next, the line scan data are processed to fill in the data missing in the circle scan data, i.e. those integration planes which do not intersect the circle scan. Essentially, with this process, space-variant filtering is applied to the linear orbit (line scan), where a redundancy function is employed to extract the 3-D Radon Data that is inaccessible from the circular orbits and discard other multiply measures 3-D Radon data. More specifically, the line scan data are processed as follows: For each angle xcex8: (1) compute all line integrals at angle xcex8; (2) compute the derivative of the line integrals; and (3) compute a 2D backprojection of the line integral derivatives. Steps (1), (2) and (3) are performed on each line such that the integration plane defined by the line and the source position does not intersect the circle scan. The last step (4) is to compute the derivative of the resultant 2D backprojection image in the horizontal direction. The above process results in reconstruction of the extended regions xcexa9u and xcexa9d as shown in FIG. 7. Further, as noted above, the method disclosed in the above-incorporated U.S. Pat. No. 5,383, 119, is used to reconstruct the region xcexa9o. The combination of the different reconstructed regions yield the extended ROT shown in FIG. 7. Advantageously, since the data acquisition process according to one embodiment of the invention utilizes the same scan paths as disclosed in U.S. Pat. No. 5,383,119 (i.e., two circular scans and one connecting line scan), one can obtain reconstruction of an extended ROT, while maintaining the same scan time and fully utilizing the detector area. In other embodiments of the invention, the data acquisition and image reconstruction may be performed using more than two circle scans. For example, FIG. 8 illustrates an extended ROT that is reconstructed from three circular scan paths, an upper scan 21, lower scan 22 and middle scan 23, and one connecting line scan 24. The volume xcexa9obetween the upper and lower scans 21, 22 is reconstructed with the data combination method disclosed in the incorporated U.S. Pat. No. 5,383, 119. Further, the portion of the cone beam image with v greater than 0 for source positions on the upper circular scan path 21, as well as the portion of the cone beam image with v less than 0 for source positions on the lower circular scan 22, is processed using the Kudo method as discussed above. FIGS. 9a, 9b and 9c illustrate detector utilization of upper, middle and lower circular scans, respectively, for data acquisition and image reconstruction of the extended ROT shown in FIG. 8. Based on the teachings herein, an extension of the current method for 4 or more circular scans is readily apparent to one of ordinary skill in the art. Although illustrative embodiments of the present invention have been described herein with reference to the accompanying drawings, it is to be understood that the invention is not limited to those precise embodiments, and that various other changes and modifications may be affected therein by one skilled in the art without departing from the scope or spirit of the invention. All such changes and modifications are intended to be included within the scope of the invention as defined by the appended claims.
051184648	claims	1. In a method of nondestructive acoustical inspection for discontinuities in solid structures including sending an ultrasound wave packet into a solid structure to be inspected, reflecting said ultrasound wave packet from a located discontinuity in the solid structure to be inspected, receiving and transducing the reflected echo of said ultrasound wave packet, and analyzing the reflected echo of said ultrasound wave packet for nondestructive interrogation of the solid structure inspected, the improved method for ultrasound inspection of solid structures through barriers in said solid structures, such as gaps in manufactured solid structures comprising the steps of: providing a first solid structure for originally receiving said interrogating ultrasound; providing a second solid structure separated from said first solid structure by a gap; flooding said gap between said first and second solid structures with a medium, said medium having a predictable speed of ultrasound transmission for said wave packet; transmitting an interrogating wave packet of ultrasound to said first solid structure, the ultrasound of said wave packet when traveling through media such as that media flooding said gap including an instantaneous standing path length within the media which is at least twice the dimension of the gap to be bridged, said interrogating wave packet of said ultrasound containing a frequency having a wavelength relative to the media in said gap to create at least one constructively interfering standing wave node between said first structure and said second solid structure within said media flooding said gap, whereby ultrasound of said frequency passes across said gap; passing said ultrasound of said frequency into said second solid structure to acoustically interrogate said second solid structure for discontinuities through reflected ultrasound of said frequency; providing a receiving transducer for receiving said reflected ultrasound of said frequency; receiving said reflected ultrasound of said frequency at said receiving transducer from the interrogated second solid structure along a path including said second solid structure, said media flooding said gap at a constructively interfering standing wave, and said first solid structure for receipt and analysis of the received reflected ultrasound. flooding said gap between said first and second solid structures with a gas, said gas having a predictable speed of ultrasound transmission for said wave packet; transmitting an interrogating wave packet of ultrasound to said first solid structure, the ultrasound of said wave packet when traveling through gas such as that gas flooding said gap including an instantaneous standing path length within the gas which is at least twice the dimension of said gap to be bridged, said interrogating wave packet of said ultrasound containing a frequency having a wavelength relative to the gas in said gap to create at least one constructively interfering standing wave node between said first solid structure and said second solid structure within said gas flooding said gap, whereby ultrasound of said frequency passes across said gap; passing said ultrasound of said frequency into said second solid structure to acoustically interrogate said second solid structure for discontinuities through reflected ultrasound of said frequency; providing a receiving transducer for receiving said reflected ultrasound of said frequency; receiving said reflected ultrasound of said frequency at said receiving transducer from the interrogated second solid structure along a path including said second solid structure, said gas flooding said gap at a constructively interfering standing wave, and said first solid structure for receipt and analysis of the received reflected ultrasound. displacing ambient gas in said gap with said flooded gas. providing a first solid structure for originally receiving said interrogating ultrasound wave packet; providing a second solid structure separated from said first structure by a gap; providing a gas in said gap; transmitting an interrogating wave packet of ultrasound to said first structure: changing said wave packets in frequency to determine when said wave packet contains a frequency which when traveling through gas such as that gas flooding said gap includes an instantaneous standing path length within the gas which is at least twice the dimension of the gap to be bridged, said interrogating wave packet of said ultrasound containing said determined frequency having a wavelength relative to the gas in said gap to create at least one constructively interfering standing wave node between said first solid structure and said second solid structure within said gas flooding said gap, whereby ultrasound of said frequency passes across said gap; passing said ultrasound of said determined frequency into said second solid structure to acoustically interrogate the second solid structure for flaws; providing a receiving transducer; receiving reflected ultrasound from the interrogated second solid structure across said gap at the constructively interfering standing wave, through the first solid structure and then to said transducer for receipt and analysis of the received ultrasound. means for providing a medium having a predictable speed of ultrasound in said gap; means for transmitting an ultrasound wave packet to said first solid structure, the ultrasound of said wave packet when traveling through medium such as that medium flooding said gap including an instantaneous standing path length within the medium which is at least twice the dimension of the gap to be bridged, said interrogating wave packet of said ultrasound containing a frequency having a wavelength relative to the medium in said gap to create at least one constructively interfering standing wave node between said first solid structure and said second solid structure within said medium flooding said gap, whereby ultrasound of said frequency passes across said gap; a receiving transducer for receiving reflected ultrasound from said transducer whereby the acoustical signal from said interrogated second solid structure across said gap at the constructively interfering standing wave, through the first solid structure and then to said transducer for receipt and analysis of the received ultrasound for flaws in said second solid structure. 2. The method of claim 1 and wherein said medium flooding said gap is a gas. 3. The method of claim 2 and wherein said gas is an inert gas. 4. The method of claim 2 and wherein said gas is helium. 5. The method of claim 1 and wherein said interrogating ultrasound is normally incident on said first solid structure. 6. The method of claim 1 and wherein said provided solid structures are steel. 7. In a method of nondestructive acoustical inspection for discontinuities in solid structures including sending an ultrasound wave packet into a solid structure to be inspected, reflecting said ultrasound wave packet from a located discontinuity in the solid structure to be inspected, receiving and transducing the reflected echo of said ultrasound wave packet, and analyzing the reflected echo of said ultrasound wave packet for nondestructive interrogation of the solid structure inspected, the improved method for ultrasound inspection of solid structures through barriers in said solid structures, such as gaps in manufactured solid structures wherein an acoustical interrogating path must pass through a first solid structure, across said gap to interrogate a secondary solid structure for said discontinuity, the inspection process comprising the steps of: 8. The process of claim 7 and wherein said transmitting step includes transmitting differing frequencies to locate those frequencies producing an optimum standing wave node in said gap. 9. The process of claim 7 and wherein said providing said gas in said gap includes the step of: 10. The process of claim 9 and wherein said flooded gas is helium. 11. The improved method for ultrasonic inspection of materials for flaws through barriers such as gaps between solid manufactured parts wherein interrogating ultrasound wave packets pass through a primary material, across said gap to interrogate a secondary material for said flaw, the improvement to said method comprising the steps of: 12. The invention of claim 11 and wherein said provided receiving transducer transmits said interrogating wave packet. 13. Apparatus for the nondestructive acoustical inspection including the ultrasonic inspection of materials through barriers such as gaps in manufactured parts between first and second solid structures wherein ultrasound passes through a first solid structure, across said gap and to interrogate a second solid structure for said flaw, the improvement comprising: 14. The apparatus of claim 13 and wherein said means for providing a medium within said gap includes means for providing a gas for displacing ambient gas in said gap. 15. The apparatus of claim 14 and wherein said provided gas comprises helium. 16. The apparatus of claim 14 and wherein said means for transmitting includes an electrically actuated ultrasound transducer, said transducer communicated to a couplant fluid between said transducer and said first material.
050948032	summary	BACKGROUND OF THE INVENTION The present invention relates to a steam generator utilized for a liquid-metal coolant reactor and more particularly, to a steam generator in which an electromagnetic pump is incorporated. In a fast breeder reactor utilizing a liquid-metal coolant, a primary coolant as a reactor coolant is exposed high levels of to radioactivity, so that it is necessary to isolate a primary cooling system from a steam generation system, and it is also necessary to carry out a heat exchange operation between the liquid metal and water during the steam generating process for supplying the steam to a turbine generator. However, in a fast breeder reactor utilizing a liquid-metal coolant, since an extremely large amount of heat is generally generated due to a chemical reaction based on the heat exchanging operation between the liquid metal and the water, it is necessary to disperse the heat generated. For this reason, a secondary cooling system is generally located between the primary cooling system and the steam generation system. A typical example of a cooling system of the liquid-metal coolant reactor of the conventional type described is shown in FIG. 8. Referring to FIG. 8, in a reactor vessel 1, disposed in a roof slab 8, are arranged a reactor core 2 and a primary cooling system comprising a primary main circulation pump 3 for circulating a liquid-metal coolant in the reactor vessel 1 for cooling the core 2, and an intermediate heat exchanger 4 for carrying out the heat exchanging operation between the primary coolant and the secondary coolant. On the other hand, a secondary cooling system comprises, as shown in FIG. 8, the intermediate heat exchanger 4, a steam generator 5 for generating steam to be supplied to a turbine generator, an electromagnetic pump 6 arranged inside the steam generator 5 for circulating the secondary coolant, and pipings 7 for connecting the equipment described above. The steam generator 5 is disposed outside the roof slab 8 which is surrounded by a wall structure of the reactor. The electromagnetic pump 6 is inserted into an upper portion of a liquid-metal outlet rising pipe 11 for sucking the liquid-metal from the lower portion of the rising pipe 11 and for feeding the same towards the intermediate heat exchanger 4. The electromagnetic pump 6 is generally provided with a stator coil (electromagnetic coil) wound around the outer periphery of an inner iron core in a spiral fashion, and an electric current is conducted to the stator coil from an external power source to thereby generate a magnetic field to cause the circulation of the liquid metal. During the conduction of the electric current and the operation of the electromagnetic pump 6, heat is generated from the stator coil. Accordingly, it is desired to effectively remove and disperse this heat during the operation thereof. For this purpose, various trials have been carried out for effectively absorbing the heat generated from the stator coil and recovering the same into the metal-liquid to suppress the energy loss during the operation of the reactor to a minimum. FIG. 9 shows one example of an electromagnetic pump proposed for the purpose of achieving the effect described above and disclosed in Japanese Utility Model Laid-open Publication No. 116701/1988. An electromagnetic pump 20 shown in FIG. 9 comprises an inner iron core 22 provided with an inner through hole 21 and an outer iron core 23 arranged concentrically with space around the outer periphery of the inner iron core 22. The space between the outer periphery of the inner iron core 22 and the inner periphery of the outer iron core 23 is formed as an annular passage 24 through which the liquid metal passes. A first stator coil (electromagnetic coil) 25 is embedded in an annular fashion in the outer peripheral surface of the inner iron core 22, and the outer surface of the first stator coil 25 is completely covered with a sealing member 26. Both of the vertical ends 26a of the sealing member 26 extend beyond the outer end portions of the inner iron core 22 and are connected with each other to be closed and thus define inner spaces 27 at both the ends of the iron core 22. These inner spaces 27 and the through hole 21 are filled with inert gas. A second stator coil (electromagnetic coil) 28 is embedded in an annular fashion in the inner peripheral surface of the outer iron core 23 and the outer surface of the second stator coil 28 is covered with a sealing member 29. The outer periphery of the outer iron core 23 is surrounded by an annular member 31 having its outer periphery supported by an electromagnetic pump supporting cylinder 29a. A plurality of bypass passages 32 are formed in the annular member 31 along the axial direction of the outer peripheral surface of the outer iron core 23 and the upper ends and the lower ends of the bypass passages 32 are provided with bypass passage inlets 33 and the bypass passage outlets 34, respectively. The electromagnetic pump 20 of the structure described above is secured to a flange member 35 which is secured to a flanged portion of the steam generator 5 used to install the electromagnetic pump 20 in the steam generator 5. The electromagnetic pump 20 operates to draw the liquid metal from a suction port 36 by the magnetic force caused by the first and second stator coils 25 and 28, and the drawn liquid metal flows upwardly in the annular passage 24 and is discharged through a discharge port 37. During this operation, the pressure at the inlet portion 33 of the bypass passage becomes larger than that at the outlet portion 34 of the bypass passage, so that a part of the liquid metal passes the bypass passages 32 and circulates around the outer iron core 23. The first stator coil 25 then generates heat, which is effectively recovered by the liquid metal to thereby suppress the temperature rise due to the heat generated by the first stator coil 25. The recovery of the generated heat by means of the circulating liquid metal possibly minimizes the energy loss in whole the steam generator. However, with the conventional steam generator of the character described above, the axial through hole 21 of the inner iron core 22 constituting the electromagnetic pump 20 and the spaces 27 defined by the sealing member 26a are closed and the spaces are filled with the inert gas, so that the first stator coil 25 embedded in the inner iron core 22 is cooled by only the liquid metal passing the circular passage 24. For this reason, the cooling effect for the first stator coil 25 of the inner iron core 22 decreases and the temperature rise of the inner iron core 22 is increased, whereby the characteristics of the electromagnetic pump 20 cannot be effectively utilized. SUMMARY OF THE INVENTION An object of the present invention is to substantially eliminate the defects and drawbacks encountered in the prior art described above and to provide a steam generator utilized for a liquid-metal coolant reactor provided with an electromagnetic pump capable of effectively cooling a stator coil means of the electromagnetic pump by the flow of liquid metal to thereby effectively reduce the energy loss of steam generator as a whole. This and other objects can be achieved according to the present invention by providing a steam generator utilized for a liquid-metal coolant reactor comprising an outer body shell of hollow cylindrical structure provided with a water inlet chamber disposed at a lower portion of the body shell, an outlet steam chamber disposed at an intermediate portion of the body shell and a liquid metal inlet portion disposed at an upper portion of the body shell, a heat transfer tube assembly arranged annularly along an inner wall of the body shell so as to connect the water inlet chamber and the outlet steam chamber, a liquid metal rising pipe assembly axially extending substantially a central portion of the body shell and arranged in a radial direction offset from an arrangement of the heat transfer tube assembly, and an electromagnetic pump means arranged at an upper portion inside the liquid metal outlet rising pipe means, the electromagnetic pump means comprising a hollow cylindrical iron core provided with a comb shaped portion at an outer peripheral surface thereof and an annular stator coil means assembled in said comb shaped portion of the cylindrical iron core, a main passage of liquid metal being formed on a side on which the stator coil means of the iron core is assembled, a cooling bypass passage being formed at substantially the central portion of the cylindrical iron core in a vertically penetrating fashion. In a preferred embodiment, the cylindrical iron core may be formed so as to have inner and outer iron core portions and the stator coil means may also be formed so as to have inner and outer stator coil portions in association with the inner and outer iron core portions. According to the steam generator of a liquid-metal coolant type reactor of the characters described above, the main flow passage of the liquid metal is formed on the side on which the stator coil of the cylindrical iron core is assembled and the cooling bypass passage is formed so as to penetrate the central portion of the steam generator so that the stator coil assembled in the cylindrical iron core can be cooled by the liquid metal circulating the outer peripheral surface of the stator coil, whereby the heat generated by the stator coil can be effectively absorbed and the excessive temperature rising of the stator coils and iron cores can be suppressed. Accordingly, the heat generation efficiency of the steam generator can be remarkably improved.
043107656	claims	1. In a sealed neutron accelerator tube having a target and a spaced replenisher section for supplying accelerator gas, an ionization section located between said target section and said replenisher section comprising: (a) means forming an ionization chamber in said tube adapted to receive accelerator gas from said replenisher section, (b) first and second cathodes spaced from one another and having opposed active surfaces exposed to the interior of said chamber, (c) anode means located at a position intermediate of said cathodes whereby in response to an applied positive voltage electrons created by field emission are transmitted between the opposed active surfaces of said cathodes and produce the emission of secondary electrons upon impacting an active cathode surface, and (d) the active surface of at least one of said cathodes being formulated by the heating in the presence of oxygen of a material having a secondary electron emission factor of at least 2, and said active surface is at least two atom layers in thickness. 2. The method of claim 1 wherein the active surfaces of both of said first and second cathodes are formulated by the heating in the presence of oxygen of a material having a secondary electron emission factor of at least 2. 3. The neutron accelerator tube of claim 1 wherein said active surface is of a thickness to be stable to ion impact. 4. The neutron accelerator of claim 1 wherein said active surface is mechanically secured to said cathode. 5. The neutron accelerator tube of claim 1 wherein said material is an alkali metal halide. 6. The neutron accelerator tube of claim 1 wherein said material is an alkaline earth metal halide.
046541845	claims	1. A method of operating a toroidal magnetic confinement device for confining a plasma, said plasma having a magnetic field B associated therewith, said magnetic field having an average magnetic pressure, B.sub.av.sup.2 /2, and said plasma having a pressure p, said pressure having an average value, p.sub.av, wherein B.sub.av.sup.2 =.intg.B.sup.2 d.tau./.intg.d.tau., and p.sub.av =.intg.pd.tau./.intg.d.tau., integration being over the plasma volume, said magnetic field and pressure defining a beta, .beta., associated with said plasma wherein .beta.=2 p.sub.av /B.sub.av.sup.2, and said plasma having a first and a second region of stability and a region of instability therebetween, said method comprising the steps of: (a) modifying the shape of said plasma until said plasma has a bean-shaped poloidal cross-section, a measure of said bean-shape being hte indentation at the inner most point on the plasma at the inboard side of the poidal cross-section; (b) maintaining said beta below the threshold for instability for operation in the first region of stability, while increasing said indentation to a critical value, said critical value being the value at which said second region of stability is accessed from said first region of stability without entering said region of instability; and (c) increasing said beta until a desired value of beta in said second region of stability is attained while maintaining said indentation at said critical value or greater. (d) reducing said indentation to a value less than said critical value and large enough to maintain said desired value of beta in said second region of stability. (a) applying a magnetic field B to said device, said field having an average magnetic pressure B.sub.av.sup.2 /2, where EQU B.sub.av.sup.2 =.intg.B.sup.2 d.tau./.intg.d.tau.; (b) forming a plasma within said device, said plasma having a bean-shaped poloidal cross-section, a measure of said bean-shape being the indentation at the innermost point on the plasma at the inboard side of the poloidal cross-section, wherein said indentation is equal to at least a critical value, said critical value being the value at which said second region of stability is accessed from said first region of stability without entering said region of instability, said plasma having a pressure p, said pressure having a average value p.sub.av, where EQU p.sub.av =.intg.pd.tau./.intg.d.tau.; (c) increasing beta, .beta., to a desired value in said second region of stability where ##EQU7## (d) reducing said indentation to a value less than said critical value and large enough to maintain said desired value of beta. 2. The method of claim 1 further comprising the step 3. The method of claim 1 wherein said plasma has plasma pressure profiles p(y) satisfying the relationship: p(y)=p.sub.o (1-y.sup.2).sup.2, where p.sub.o is a constant, y=.psi./.DELTA..psi. and 2.pi..DELTA..psi. is the poloidal flux in the plasma. 4. The method of claim 1 wherein said plasma has a safety factor profile q(y) satisfying the relationship: ##EQU5## where q.sub.i are constants, y=.psi./.DELTA..psi. and 2.pi..DELTA..psi. is the poloidal flux in the plasma. 5. The method of claim 1 wherein the shape of said plasma cross-section is given by EQU x(t)=x+.rho. cos .gamma., EQU z(t)=E.rho. sin .gamma. 6. The method of claim 1 wherein beta is adjusted by varying at least one of p.sub.av and B.sub.av.sup.2. 7. The method of claim 1 wherein said plasma has a plasma pressure profile p(y) and safety factor profile q(y) satisfying the relationships: ##EQU6## where p.sub.o, q.sub.i are constants, y=.psi./.DELTA..sub..psi., 2.pi..DELTA..sub..psi. is the poloidal flux in the plasma. 8. The method of claim 7 wherein said device has an aspect ratio of 4 and said critical value is about 0.304. 9. The method of claim 7 wherein said device has an aspect ratio of 7 and said critical value is about 0.33. 10. The method of claim 7 wherein said device has an aspect ratio of 10 and said critical value is about 0.35. 11. The method of claim 7 wherein said device is a tokamak. 12. A method of operating a toroidal magnetic confinement device, for confining a plasma, said plasma having a first and a second region of stability and a region of instability therebetween, said method comprising the steps of: 13. The method of claim 12 further comprising the step 14. The method of claim 13 wherein said device is a tokamak. 15. The method of claim 14 wherein said plasma has pressure profiles p(y) satisfying the relationship: EQU p(y)=p.sub.o (1-y.sup.2).sup.2, 16. The method of claim 15 wherein said plasma has a safety factor profile q(y) satisfying the relationship: ##EQU8## where q.sub.i are constant. 17. The method of claim 16 wherein the shape of said plasma cross-section is given by EQU x(t)=x+.rho. cos .gamma., EQU z(t)=E.rho. sin .gamma., 18. The method of claim 12 wherein said bean shaped cross-section is formed by energizing a pusher coil located at the inner major radius side of the plasma. 19. The method of claim 12 wherein beta is established by varying at least one of p.sub.av and B.sub.av.sup.2. 20. The method of claim 1 wherein said bean-shaped cross-section is formed by energizing a pusher coil located at the inner major radius side of the plasma. 21. The method of claim 18 wherein the pusher coil is located in the vicinity of the indentation and is energized by an external current source. 22. The method of claim 20 wherein the pusher coil is located in the vicinity of the indentation and is energized by an external current source. 23. The method of claim 21 wherein the magnitude of the current through the pusher coil varies with time. 24. The method of claim 22 wherein the magnitude of the current through the pusher coil varies with time.
054935903	abstract	A critical power enhancement system is provided for a pressurized fuel channel type nuclear reactor comprising a plurality of fuel bundles contained in a fuel channel and containing a plurality of fuel elements horizontally oriented within the fuel channel. The system comprises at least one appendage strategically located on each of certain fuel elements along its length and projecting outwardly from the surface of the fuel element. The appendages generate turbulence in the coolant flowing at locations along the length of the fuel bundle, where the critical heat flux is most likely to occur. The presence of the appendages suppress the occurrence of the critical heat flux in the fuel bundle thereby increasing the safety limit on the maximum power that can be produced by the reactor.
	abstract	Disclosed is an electrostatic chuck with a temperature sensing unit, exposure equipment having the electrostatic chuck, and a method of detecting temperature on photomask surfaces. The temperature sensing unit and method of detecting temperature may include obtaining reflectance of a photomask using a multi-wavelength interferometer and determining a temperature on the photomask based on the reflectance.
	summary
	description	This application is a continuation of U.S. patent application Ser. No. 15/427,210, filed Feb. 8, 2017, now U.S. Pat. No. 10,648,958, which is a divisional application of U.S. application Ser. No. 13/725,073, filed Dec. 21, 2012, which claims the benefit of U.S. Provisional Application Ser. No. 61/578,431, filed Dec. 21, 2011, each of which is incorporated herein by reference in its entirety, and to which applications we claim priority under 35 USC § 121. This invention was made with government support under Grant Number HR0011-10-3-0002, awarded by the U.S. Department of Defense, Defense Advanced Research Projects Agency, Microsystems Technology Office. The government has certain rights in the invention. The present disclosure provides an interconnected corrugated carbon-based network and an inexpensive process for making, patterning, and tuning the electrical, physical and electrochemical properties of the interconnected corrugated carbon-based network. In the pursuit of producing high quality bulk carbon-based devices such as organic sensors, a variety of syntheses now incorporate graphite oxide (GO) as a precursor for the generation of large scale carbon-based materials. Inexpensive methods for producing large quantities of GO from the oxidation of graphitic powders are now available. In addition, the water dispersibility of GO combined with inexpensive production methods make GO an ideal starting material for producing carbon-based devices. In particular, GO has water dispersible properties. Unfortunately, the same oxygen species that give GO its water dispersible properties also create defects in its electronic structure, and as a result, GO is an electrically insulating material. Therefore, the development of device grade carbon-based films with superior electronic properties requires the removal of these oxygen species, re-establishment of a conjugated carbon network, as well as a method for controllably patterning carbon-based device features. Methods for reducing graphite oxide have included chemical reduction via hydrazine, hydrazine derivatives, or other reducing agents, high temperature annealing under chemical reducing gases and/or inert atmospheres, solvothermal reduction, a combination of chemical and thermal reduction methods, flash reduction, and most recently, laser reduction of GO. Although several of these methods have demonstrated relatively high quality graphite oxide reduction, many have been limited by expensive equipment, high annealing temperatures and nitrogen impurities in the final product. As a result, of these difficulties, a combination of properties that includes high surface area and high electrical conductivity in an expanded interconnected carbon network has remained elusive. In addition, large scale film patterning via an all-encompassing step for both GO reduction and patterning has proven difficult and has typically been dependent on photo-masks to provide the most basic of patterns. Therefore, what is needed is an inexpensive process for making and patterning an interconnected corrugated carbon-based network having a high surface area with highly tunable electrical conductivity and electrochemical properties. The present disclosure provides a method of producing an interconnected corrugated carbon-based network. The interconnected corrugated carbon-based network produced has a combination of properties that includes high surface area and high electrical conductivity in an expanded network of interconnected carbon layers. In one embodiment, the method produces a patterned interconnected corrugated carbon-based network. In that particular embodiment, an initial step receives a substrate having a carbon-based oxide film. Once the substrate is received, a next step involves generating a light beam having a power density sufficient to reduce portions of the carbon-based oxide film to an interconnected corrugated carbon-based network. Another step involves directing the light beam across the carbon-based oxide film in a predetermined pattern via a computerized control system while adjusting the power density of the light beam via the computerized control system according to predetermined power density data associated with the predetermined pattern. In one embodiment, the substrate is a disc-shaped, digital versatile disc (DVD) sized thin plastic sheet removably adhered to a DVD sized plate that includes a DVD centering hole. The DVD sized plate carrying the disc-shaped substrate is loadable into a direct-to-disc labeling enabled optical disc drive. A software program executed by the computerized control system reads data that defines the predetermined pattern. The computerized control system directs a laser beam generated by the optical disc drive onto the disc-shaped substrate, thereby reducing portions of the carbon-based oxide film to an electrically conductive interconnected corrugated carbon-based network that matches shapes, dimensions, and conductance levels dictated by the data of the predetermined pattern. Those skilled in the art will appreciate the scope of the disclosure and realize additional aspects thereof after reading the following detailed description in association with the accompanying drawings. All publications, patents, and patent applications mentioned in this specification are herein incorporated by reference to the same extent as if each individual publication, patent, or patent application was specifically and individually indicated to be incorporated by reference. While preferred embodiments of the present invention have been shown and described herein, it will be obvious to those skilled in the art that such embodiments are provided by way of example only. Numerous variations, changes, and substitutions will now occur to those skilled in the art without departing from the invention. It should be understood that various alternatives to the embodiments of the invention described herein may be employed in practicing the invention. It is intended that the following claims define the scope of the invention and that methods and structures within the scope of these claims and their equivalents be covered thereby. The embodiments set forth below represent the necessary information to enable those skilled in the art to practice the disclosure and illustrate the best mode of practicing the disclosure. Upon reading the following description in light of the accompanying drawings, those skilled in the art will understand the concepts of the disclosure and will recognize applications of these concepts not particularly addressed herein. It should be understood that these concepts and applications fall within the scope of the disclosure and the accompanying claims. The present disclosure provides an inexpensive process for making and patterning an interconnected corrugated carbon-based network having stringent requirements for a high surface area with highly tunable electrical conductivity and electrochemical properties. The embodiments described herein not only meet these stringent requirements, but provide direct control over the conductivity and patterning of interconnected corrugated carbon-based networks while creating flexible electronic devices in a single step process. Moreover, the production of these interconnected corrugated carbon-based networks does not require reducing agents, or expensive equipment. The simple direct fabrication of interconnected corrugated carbon-based networks on flexible substrates therefore simplifies the development of lightweight electronic devices. The interconnected corrugated carbon-based networks can be synthesized on various substrates, such as plastic, metal, and glass. Herein an all-organic NO2 gas sensor, a fast redox active electrode, and a scaffold for the direct growth of platinum (Pt) nanoparticles are demonstrated. In at least one embodiment, the interconnected corrugated carbon-based networks are conducting films produced using a common and inexpensive infrared laser that fits inside a compact disc/digital versatile disc (CD/DVD) optical drive unit that provides a direct-to-disc label writing function. LightScribe (Registered Trademark of Hewlett Packard Corporation) and LabelFlash (Registered Trademark of Yamaha Corporation) are exemplary direct-to-disc labeling technologies that pattern text and graphics onto the surface of a CD/DVD disc. LightScribe DVD drives are commercially available for around $20 and the LightScribing process is controlled using a standard desktop computer. FIG. 1 depicts the label side of a standard direct-to-disc labeling type CD/DVD disc 10 that includes a label area 12 and a clamping area 14 that surrounds a centering hole 16. A dye film 18 covers the label area 12 and is sensitive to laser energy that is typically directed onto the label area 12 to produce a permanent visible image that may comprise graphics 20 and text 22. A position tracking indicia 24 is usable by an optical disc drive (not shown) to accurately locate an absolute angular position of the CD/DVD disc 10 within the optical disc drive so that the graphics 20 and/or text 22 can be re-written to provide increased contrast. Moreover, the position tracking indicia 24 is usable by the optical disc drive to allow additional graphics and/or text to be written without undesirably overwriting the graphics 20 and/or text 22. FIG. 2 is a schematic of a prior art direct-to-disc labeling type optical disc drive system 26. In this exemplary case, the CD/DVD disc 10 is depicted in cross-section and loaded onto a spindle assembly 28 that is driven by a CD/DVD spindle motor 30. The label area 12 is shown facing a laser assembly 32 that includes a label writer laser (LWL) 34, a lens 36, and a focus actuator 38. The LWL 34 is typically a laser diode. Exemplary specifications for the LWL 34 includes a maximum pulse optical power of 350 mW at 780 nm emission and a maximum pulse output power of 300 mW at 660 nm emission. A laser beam 40 emitted by the LWL 34 is focused by the lens 36 that is alternately translated towards and away from the LWL 34 by the focus actuator 38 in order to maintain focus of the laser beam 40 onto the label area 12 of the CD/DVD disc 10. The laser beam 40 is typically focused to a diameter that ranges from around 0.7 μm to around 1 μm. The laser assembly 32 is responsive to a control system 42 that provides control signals 44 through an optical drive interface (ODI) 46. The control system 42 further includes a central processor unit (CPU) 48 and a memory 50. Label image data (LID) having information needed to realize a permanent image to be written onto the label area 12 of the CD/DVD disc 10 is processed by the CPU 48, which in turn provides an LID stream signal 52 that pulses the LWL 34 on and off to heat the dye film 18 to realize the image defined by the LID. The CPU 48 also processes the LID through the ODI 46 to provide a position control signal 54 to a radial actuator 56 that translates the laser assembly 32 in relation to the label area 12 in response to position information contained in the LID. In some versions of the present embodiments, the optical disc drive system 26 monitors the focus of the laser beam 40 with an optical receiver (not shown), so that the ODI 46 can generate a focus control signal 58 for the focus actuator 38. The ODI 46 also provides a motor control signal 60 for the CD/DVD spindle motor 30 that maintains an appropriate rotation speed of the CD/DVD disc 10 while a label writing process is ongoing. In some versions of the optical disc drive system 26 the LWL 34 is used exclusively for label writing directly to the label area 12 of the CD/DVD disc 10 and a separate laser diode (not shown) is used to write and/or read data to/from a data side 62 of the CD/DVD disc 10. In other versions of the optical disc drive system 26, the LWL 34 is used for label writing and data reading and/or writing. When the LWL 34 is used for data reading and/or writing, the CD/DVD disc 10 is flipped over to expose the data side 62 of the CD/DVD disc 10 to the laser beam 40. In versions wherein the LWL 34 is also used as a data read/write laser, the laser assembly 32 includes optical pick-up components (not shown) such as a beam splitter and at least one optical receiver. The output power of the LWL 34 is typically around 3 mW during data read operations. In order to use the optical disc drive system 26 to realize an inexpensive process for making and patterning an interconnected corrugated carbon-based network having a high surface area with highly tunable electrical conductivity and electrochemical properties, a carbon-based film is substituted for the dye film 18 (FIG. 1). In one embodiment, graphite oxide (GO) is synthesized from high purity graphite powder using a modified Hummer's method. Dispersions of GO in water (3.7 mg/mL) are then used to make GO films on various substrates. Exemplary substrates include but are not limited to polyethylene terephthalate (PET), nitrocellulose membrane (with 0.4 μm pore size), aluminum foil, carbonized aluminum, copper foil, and regular copier paper. Referring to FIG. 3, a process 100 begins with providing graphite powder 64. The graphite powder 64 undergoes an oxidation reaction using the modified Hummer's method to become GO 66 (step 102). However, it is to be understood that other oxidation methods for producing GO are available and such methods are within the scope of the present disclosure. An exfoliation procedure produces exfoliated GO 68 (step 104). The exfoliation procedure may be accomplished via ultrasonication. It is to be understood that the exfoliated GO 68 results from a partial exfoliation and not a complete exfoliation to a single layer of GO. The partial exfoliation is used to create a high accessible surface area that enables a fast redox response which enables a fast sensor response. Additionally, the partial exfoliation of GO 68 provides the high surface area for growing metal nanoparticles that could then be used in catalysis. A substrate 70 carries a GO film 72 that is produced by a deposition procedure that deposits the exfoliated GO 68 onto the substrate 70 (step 106). In at least some embodiments, a GO film 72 is made by either drop-casting or vacuum filtering GO dispersions onto the substrate 70 that is the size of a CD/DVD disc. The GO film 72 is typically allowed to dry for 24 hours under ambient conditions. However, controlling conditions to expose the GO film 72 to a relatively lower humidity and relatively higher temperature will dry the GO film 72 relatively quickly. The term GO herein refers to graphite oxide. Referring to FIG. 4, individual ones of the GO film(s) 72 are then affixed to a substrate carrier 74, which has dimensions similar to the CD/DVD disc 10 (FIG. 1) (step 108). The substrate carrier 74 carrying the substrate 70 with the GO film 72 is loaded into the optical disc drive system 26 (FIG. 2) such that the GO film 72 faces the LWL 34 for laser treatment (step 110). In this way, the present embodiments use the GO film 72 in place of the dye film 18 (FIG. 1). It is to be understood that the substrate carrier 74 can be a rigid or semi-rigid disc onto which the GO film 72 can be fabricated directly. In that case, the substrate carrier 74 replaces the function of the substrate 70. Images 76 for realizing electrical components 78 are patterned in concentric circles, moving outward from the center of the substrate carrier 74 (step 112). The laser irradiation process results in the removal of oxygen species and the reestablishment of sp2 carbons. This causes a change in the conductivity of the GO film 72 with a typical resistance of >20 MΩ/sq to become a relatively highly conducting plurality of expanded and interconnected carbon layers that make up an interconnected corrugated carbon-based network 80. The number of times the GO film 72 is laser treated results in a significant and controllable change in the conductivity of the interconnected corrugated carbon-based network 80. The interconnected corrugated carbon-based network 80 has a combination of properties that include high surface area and high electrical conductivity in an expanded interconnected network of carbon layers. In one embodiment the plurality of expanded and interconnected carbon layers has a surface area of greater than 1400 m2/g. In another embodiment, the plurality of expanded and interconnected carbon layers has a surface area of greater than 1500 m2/g. In yet another embodiment, the surface area is around about 1520 m2/g. In one embodiment, the plurality of expanded and interconnected carbon layers yields an electrical conductivity that is greater than about 1500 S/m. In another embodiment, the plurality of expanded and interconnected carbon layers yields an electrical conductivity that is greater than about 1600 S/m. In yet another embodiment, the plurality of expanded and interconnected carbon layers yields an electrical conductivity of around about 1650 S/m. In still another embodiment, the plurality of expanded and interconnected carbon layers yields an electrical conductivity that is greater than about 1700 S/m. In yet one more embodiment, the plurality of expanded and interconnected carbon layers yields an electrical conductivity of around about 1738 S/m. Moreover, in one embodiment, the plurality of expanded and interconnected carbon layers yields an electrical conductivity that is greater than about 1700 S/m and a surface area that is greater than about 1500 m2/g. In another embodiment, the plurality of expanded and interconnected carbon layers yields an electrical conductivity of around about 1650 S/m and a surface area of around about 1520 m2/g. The electrical components 78 comprising electrodes 82 used in the fabrication of a device 84 are laser irradiated 6 times before reaching the relatively high conductivity of around about 1738 S/m. The laser irradiation process takes about 20 minutes per cycle. Afterwards, the substrate 70 carrying the interconnected corrugated carbon-based network 80 and any remaining GO film 72 is removed from the substrate carrier 74 (step 114). Next, the interconnected corrugated carbon-based network 80 is fabricated into the electrical components 78 that make up the device 84 (step 116). In this exemplary case, portions of the interconnected corrugated carbon-based network 80 on the substrate 70 are cut into rectangular sections to make the electrical components 78, which include the electrodes 82 formed from the interconnected corrugated carbon-based network 80. The interconnected corrugated carbon-based network 80 possesses a very low oxygen content of only 3.5%. In other embodiments, the oxygen content of the expanded and interconnected carbon layers ranges from around about 1% to around about 5%. FIG. 5 is a line drawing of a sample of the interconnected corrugated carbon-based network 80, which is made up of the plurality of expanded and interconnected carbon layers that include corrugated carbon layers such as a single corrugated carbon sheet 86. In one embodiment, each of the expanded and interconnected carbon layers comprises at least one corrugated carbon sheet that is one atom thick. In another embodiment, each of the expanded and interconnected carbon layers comprises a plurality of corrugated carbon sheets that are each one atom thick. The thickness of the interconnected corrugated carbon-based network 80, as measured from cross-sectional scanning electron microscopy (SEM) and profilometry, was found to be around about 7.6 μm. In one embodiment, a range of thickness of the plurality of expanded and interconnected carbon layers making up the interconnected corrugated carbon-based network 80 is from around 7 μm to 8 μm. As an illustration of the diversity in image patterning that is possible, a complex image formed by the direct laser reduction of GO is shown in FIGS. 6A and 6B. FIG. 6A is an artwork image of a man's head covered with circuits. FIG. 6B is a photograph of a GO film after the artwork image of FIG. 6A is directly patterned on the GO film using the laser scribing technique of the present disclosure. Essentially, any part of the GO film that comes in direct contact with the 780 nm infrared laser is effectively reduced to an interconnected corrugated carbon-based network, with the amount of reduction being controlled by the laser intensity; a factor that is determined by power density of the laser beam impinging on the GO film. The resulting image of FIG. 6B is an effective print of the original image of FIG. 6A. However, in this case the image of FIG. 6B is made up of various reductions of the GO film. As expected, the darkest black areas indicate exposure to the strongest laser intensities, while the lighter gray areas are only partially reduced. Since different grayscale levels directly correlate with the laser's intensity, it is possible to tune the electrical properties of the generated interconnected corrugated carbon-based network over five to seven orders of magnitude in sheet resistance (Ω/sq) by simply changing the grayscale level used during the patterning process. As illustrated in FIG. 7, there is a clear relationship between sheet resistance, grayscale level and the number of times the GO film is laser irradiated. Control over conductivity from a completely insulating GO film, with a typical sheet resistance value of >20 MΩ/sq, to a conducting interconnected corrugated carbon-based network that registers a sheet resistance value of approximately 80 Ω/sq, which translates to a conductivity of ˜1650 S/m, is possible. This method is sensitive enough to differentiate between similar grayscale levels as shown in the graph of FIG. 7, where the sheet resistance varies significantly with only a small variation in grayscale level. In addition, the number of times a GO film is laser treated results in a significant and controllable change in sheet resistance. Each additional laser treatment lowers the sheet resistance as seen in FIG. 7, where a film is laser irradiated once (black squares), twice (circles) and three times (triangles) with respect to the grayscale level. Therefore, the film's sheet resistance is tunable both by controlling the grayscale level used and the number of times the film is reduced by the laser, a property that has so far been difficult to control through other methods. Scanning electron microscope (SEM) techniques are usable to understand the effects a low energy infrared laser has on the structural properties of GO film by comparing the morphological differences between an interconnected corrugated carbon-based network and untreated graphite oxide GO film. FIG. 8A is an SEM image that illustrates the infrared laser's effect on GO film prior to laser treatment on the right side of the image in contrast to an aligned, interconnected corrugated carbon-based network on the left side of the image that occurs after being reduced with the infrared laser. The image not only gives a clear definition between the interconnected corrugated carbon-based network and untreated GO regions, but also demonstrates the level of precision possible when using this method as a means to pattern and reduce GO. The regions of interconnected corrugated carbon-based network, which result from the laser treatment, can be further analyzed through cross-sectional SEM. FIG. 8B is an SEM image showing a cross-sectional view of a free standing film of laser treated and untreated GO film, which shows a significant difference between GO film thicknesses. As indicated by the white brackets in FIG. 8B, an interconnected corrugated carbon-based network increases in thickness by approximately 10 times in comparison to that of untreated GO film. Moreover, a range of thickness of the plurality of expanded and interconnected carbon layers is from around 7 μm to around 8 μm. In one embodiment, an average thickness of the plurality of expanded and interconnected carbon layers is around 7.6 μm. The increased thickness stems from rapid degassing of gases generated and released during laser treatment, similar to thermal shock, which effectively causes the reduced GO to expand and exfoliate as these gases rapidly pass through the GO film. FIG. 8C is an SEM image showing a cross-sectional view of a single interconnected corrugated carbon-based network, which shows an expanded structure that is a characteristic of the interconnected corrugated carbon-based network of the present disclosure. FIG. 8D is an SEM image showing a greater magnification of a selected area within the corrugated carbon-based network in FIG. 8C. The SEM image of FIG. 8D allows the thickness of the plurality of expanded and interconnected carbon layers to be calculated to be between 5-10 nm. However, the number of carbon layers in the plurality of expanded and interconnected carbon layers making up the interconnected corrugated carbon-based network is above 100. In another embodiment the number of carbon layers in the plurality of expanded and interconnected carbon layers is greater than 1000. In yet another embodiment the number of carbon layers in the plurality of expanded and interconnected carbon layers is greater than 10,000. In still another embodiment, the number of carbon layers in the plurality of expanded and interconnected carbon layers is greater than 100,000. The SEM analysis shows that although an infrared laser emission is only marginally absorbed by GO, enough power and focus (i.e., power density) can cause sufficient thermal energy to efficiently reduce, deoxygenate, expand, and exfoliate the GO film. Moreover, the surface area of the interconnected corrugated carbon-based network is greater than about 1500 m2/g. Since each of the carbon layers have a theoretical surface area of 2630 m2/g, a surface greater than 1500 m2/g indicates that almost all surfaces of the carbon layers are accessible. The interconnected corrugated carbon-based network has an electrical conductivity that is greater than 17 S/cm. The interconnected corrugated carbon-based network forms when some wavelength of light hits the surface of the GO, and is then absorbed to practically immediately convert to heat, which liberates carbon dioxide (CO2). Exemplary light sources include but are not limited to a 780 nm laser, a green laser, and a flash lamp. The light beam emission of the light sources may range from near infrared to ultraviolet wavelengths. The typical carbon content of the interconnected corrugated carbon-based network is greater than 97% with less than 3% oxygen remaining. Some samples of the interconnected corrugated carbon-based network are greater than 99% carbon even though the laser reduction process is conducted in the air. FIG. 9 compares a powder X-ray diffraction (XRD) pattern of the corrugated carbon-based network with both graphite and graphite oxide diffraction patterns. A typical XRD pattern for graphite, shown in FIG. 9 trace A, displays the characteristic peak of 2θ=27.8° with a d-spacing of 3.20 Å. An XRD pattern (FIG. 9, trace B) for GO, on the other hand, exhibits a single peak of 2θ=10.76°, which corresponds to an interlayer d-spacing of 8.22 Å. The increased d-spacing in GO is due to the oxygen containing functional groups in graphite oxide sheets, which tend to trap water molecules between the basal planes, causing the sheets to expand and separate. The XRD pattern of the corrugated carbon-based network (FIG. 9, trace C) shows the presence of both GO (10.76° 2θ) and a broad graphitic peak at 25.97° 2θ associated with a d-spacing of 3.43 Å, (FIG. 10). The GO presence in the corrugated carbon-based network is expected since the laser has a desirable penetration depth, which results in the reduction of only the top portion of the film with the bottom layer being unaffected by the laser. The small presence of GO is more prominent in thicker films, but begins to diminish in thinner films. In addition, one can also observe a partially obstructed peak at 26.66° 2θ, which shows a similar intensity to the broad 25.97° 2θ peak. Both of these peaks are considered graphitic peaks, which are associated to two different lattice spacing between basal planes. It has been previously shown that the immobilization of carbon nanotubes (CNTs) on glassy carbon electrodes will result in a thin CNT film, which directly affects the voltammetric behavior of the CNT modified electrodes. In a ferri/ferrocyanide redox couple, the voltammetric current measured at the CNT modified electrode will likely have two types of contributions. The thin layer effect is a significant contributor to the voltammetric current. The thin layer effect stems from the oxidation of ferrocyanide ions, which are trapped between the nanotubes. The other contribution results from the semi-infinite diffusion of ferrocyanide towards the planar electrode surface. Unfortunately, the mechanistic information is not easily de-convoluted and requires knowledge of the film thickness. In contrast, no thin layer effect is observed in association with the interconnected corrugated carbon-based network of the present disclosure. FIG. 10 is a plot of log10 of peak current versus log10 of an applied voltammetric scan rate. In this case, no thin layer effect is observed since the plot has a consistent slope of 0.53 and is linear. The slope of 0.53 is relatively close to theoretical values calculated using a semi-infinite diffusion model governed by the Randles-Sevcik equation: i p = 0 . 3 ⁢ 443 ⁢ A ⁢ ⁢ C o * ⁢ D o ⁢ v ⁡ ( nF ) 3 R ⁢ T Raman spectroscopy is used to characterize and compare the structural changes induced by laser treating GO film. FIGS. 11A-11E are graphs related to Raman spectroscopic analysis. As can be seen in FIG. 11A, characteristic D, G, 2D and S3 peaks are observed in both GO and the interconnected corrugated carbon-based network. The presence of the D band in both spectra suggests that carbon sp3 centers still exist after reduction. Interestingly, the spectrum of the interconnected corrugated carbon-based network shows a slight increase in the D band peak at ˜1350 cm−1. This unexpected increase is due to a larger presence of structural edge defects and indicates an overall increase in the amount of smaller graphite domains. The result is consistent with SEM analysis, where the generation of exfoliated accordion-like graphitic regions (FIG. 5) caused by the laser treatment creates a large number of edges. However the D band also shows a significant overall peak narrowing, suggesting a decrease in the types of defects in the interconnected corrugated carbon-based network. The G band experiences a narrowing and a decrease in peak intensity as well as a peak shift from 1585 to 1579 cm−1. These results are consistent with the re-establishment of sp2 carbons and a decrease in structural defects within the basal planes. The overall changes in the G band indicate a transition from an amorphous carbon state to a more crystalline carbon state. In addition, a prominent and shifted 2D peak from around about 2730 to around about 2688 cm−1 is seen after GO is treated with the infrared laser, indicating a considerable reduction of the GO film and strongly points to the presence of a few-layer interconnected graphite structure. In one embodiment, the 2D Raman peak for the interconnected corrugated carbon-based network shifts from around about 2700 cm−1 to around about 2600 cm−1 after the interconnected corrugated carbon-based network is reduced from a carbon-based oxide. Moreover, as a result of lattice disorder, the combination of D-G generates an S3 second order peak, which appears at ˜2927 cm−1 and, as expected, diminishes with decreasing disorder after infrared laser treatment. In some embodiments, the plurality of expanded and interconnected carbon layers has a range of Raman spectroscopy S3 second order peak that ranges from around about 2920 cm−1 to around about 2930 cm−1. The Raman analysis demonstrates the effectiveness of treating GO with an infrared laser as a means to effectively and controllably produce the interconnected corrugated carbon-based network. X-ray photoelectron spectroscopy (XPS) was employed to correlate the effects of laser irradiation on the oxygen functionalities and to monitor the structural changes on the GO film. Comparing the carbon to oxygen (C/O) ratios between GO and the interconnected corrugated carbon-based network provides an effective measurement of the extent of reduction achieved using a simple low energy infrared laser. FIG. 11B illustrates the significant disparity between the C/O ratios before and after laser treatment of the GO films. Prior to laser reduction, typical GO films have a C/O ratio of approximately 2.6:1, corresponding to a carbon/oxygen content of ˜72% and 38%. On the other hand, the interconnected corrugated carbon-based network has an enhanced carbon content of 96.5% and a diminished oxygen content of 3.5%, giving an overall C/O ratio of 27.8:1. Since the laser reduction process takes place under ambient conditions, it is postulated that some of the oxygen present in the interconnected corrugated carbon-based network film is a result of the film having a static interaction with oxygen found in the environment. FIG. 11C shows that the C1s XPS spectrum of GO displays two broad peaks, which can be resolved into three different carbon components corresponding to the functional groups typically found on the GO surface, in addition to a small π to π* peak at 290.4 eV. These functional groups include carboxyl, sp3 carbons in the form of epoxide and hydroxyl, and sp2 carbons, which are associated with the following binding energies: approximately 288.1, 286.8 and 284.6 eV, respectively. FIG. 11D shows expected results, in that the large degree of oxidation in GO results in various oxygen components in the GO C1s XPS spectrum. These results are in contrast to the interconnected corrugated carbon-based network, which shows a significant decrease in oxygen containing functional groups and an overall increase in the C—C sp2 carbon peak. This points to an efficient deoxygenating process as well as the re-establishment of C═C bonds in the interconnected corrugated carbon-based network. These results are consistent with the Raman analysis. Thus, an infrared laser such as LWL 34 (FIG. 2) is powerful enough to remove a majority of the oxygen functional groups, as is evident in the XPS spectrum of the interconnected corrugated carbon-based network, which only shows a small disorder peak and a peak at 287.6 eV. The latter corresponds to the presence of sp3 type carbons suggesting that a small amount of carboxyl groups remain in the final product. In addition, the presence of a π to π* satellite peak at ˜290.7 eV indicates that delocalized π conjugation is significantly stronger in the interconnected corrugated carbon-based network as this peak is miniscule in the GO XPS spectrum. The appearance of the delocalized π peak is a clear indication that conjugation in the GO film is restored during the laser reduction process and adds support that an sp2 carbon network has been re-established. The decreased intensity of the oxygen containing functional groups, the dominating C═C bond peak and the presence of the delocalized π conjugation all indicate that a low energy infrared laser is an effective tool in the generation of the interconnected corrugated carbon-based network. FIG. 11E depicts UV-visible light absorbance spectra of GO shown in black. The inset shows a magnified view of the boxed area showing the absorbance of GO with respect to a 780 nm infrared laser in the 650 to 850 nm region. The future development of multifunctional flexible electronics such as roll-up displays, photovoltaic cells, and even wearable devices presents new challenges for designing and fabricating lightweight, flexible energy storage devices. Embodiments of the present disclosure also include other types of electrical and electronic devices. For example, FIG. 12A shows a set of interdigitated electrodes with dimensions of 6 mm×6 mm, spaced at ˜500 μm, that are directly patterned onto a thin film of GO. Prior to being patterned, the GO film was deposited on a thin flexible substrate, polyethylene terephthalate (PET), in order to fabricate a set of electrodes that are mechanically flexible. The top arrow points to the region of the interconnected corrugated carbon-based network that makes up the black interdigitated electrodes, while the bottom arrow points to the un-reduced golden colored GO film. Since the electrodes are directly patterned onto the GO film on a flexible substrate, the need for post-processing such as transferring the film to a new substrate is unnecessary. Although, if desired, a peel and stick method could be used to selectively lift-off the black interdigitated electrodes made of interconnected corrugated carbon-based networks with e.g. polydimethylsiloxane (PDMS) and transfer it onto other types of substrates (FIG. 12B). The simplicity of this method allows substantial control over pattern dimensions, substrate selectivity and electrical properties of the interconnected corrugated carbon-based network by controlling the laser intensity and thereby the amount of reduction in each film. These interdigitated electrodes can, in turn, be used as an all-organic flexible gas sensor for the detection of NO2. FIG. 13 shows the sensor response for a patterned flexible set of interdigitated electrodes made of interconnected corrugated carbon-based networks that are exposed to 20 ppm of NO2 in dry air. This sensor was fabricated by patterning interconnected corrugated carbon-based networks to fabricate the active electrode and marginally reducing the area in between the electrodes to have a consistent sheet resistance of ˜7775 ohms/sq. In this way, it is possible to bypass the use of metal electrodes and directly pattern both the electrode and the sensing material on the flexible substrate simultaneously. The plot relates NO2 gas exposure to R/R0, where R0 is the sheet resistance at the initial state and R is the resistance of the interconnected corrugated carbon-based networks film after exposure to the gas. The film was exposed to NO2 gas for 10 min followed immediately by purging with air for another 10 min. This process was then repeated nine more times for a total of 200 min. Even with a slightly lower sensitivity than more sophisticated and optimized sensors, the un-optimized sensor made up of interconnected corrugated carbon-based networks still shows good, reversible sensing for NO2 and its easy fabrication makes it quite advantageous for these systems. The sensor made up of interconnected corrugated carbon-based networks for NO2 holds promise for improving the fabrication of all-organic flexible sensor devices, at low cost by using inexpensive starting materials directly patterned with an inexpensive laser. The high conductivity and increased surface area resulting from the plurality of expanded and interconnected carbon layers, makes interconnected corrugated carbon-based networks a viable candidate for use as a heterogeneous catalyst support for metal nanoparticles. In particular, the direct growth of Pt nanoparticles on interconnected corrugated carbon-based networks could aid in the improvement of methanol based fuel cells, which have shown enhanced device performance from large surface area and conducting carbon-based scaffolds. This disclosure demonstrates that an interconnected corrugated carbon-based network is a viable scaffold for the controllable growth of Pt nanoparticles. By electrochemically reducing 1 mM of K2PtCl4 with 0.5 M H2SO4 at −0.25 V for different periods of time, it is possible to actively control the Pt particle size that is electrodeposited on the interconnected corrugated carbon-based network film. FIGS. 14A-14D shows scanning electron microscopy images illustrating the growth of Pt nanoparticles with respect to electrodeposition times corresponding to 0, 15, 60 and 120 seconds. As expected, there are no Pt particles present at 0 seconds of electrodeposition (FIG. 14A), but small Pt nanoparticles are clearly visible after just 15 seconds (FIG. 14B) with nanoparticle sizes ranging from 10-50 nm (FIG. 14B, inset). After 60 seconds of electrodeposition, larger Pt nanoparticles grow with particle sizes averaging 100 to 150 nm (FIG. 14C). Finally, after 120 seconds, 200 to 300 nm particles are found evenly distributed across the surface of the interconnected corrugated carbon-based networks (FIG. 14D). The active growth of Pt nanoparticles at controllable diameters on interconnected corrugated carbon-based networks could make a potentially useful hybrid material for applications that require metal nanoparticles, such as methanol fuel cells and gas phase catalysts. Moreover, if palladium (Pd) is deposited a sensor made of an interconnected corrugated carbon-based network could be used for sensors that detect hydrogen or for catalysis such as Suzuki coupling or Heck coupling. Carbon electrodes have attracted tremendous interest for various electrochemical applications because of their wide potential window and good electrocatalytic activity for many redox reactions. Given its high surface area and flexibility and the fact that it is an all-carbon electrode, interconnected corrugated carbon-based networks could revolutionize electrochemical systems by making miniaturized and fully flexible devices. Here, understanding the electrochemical properties of interconnected corrugated carbon-based networks is highly beneficial to determining its potential for electrochemical applications. Recently, graphene's electrocatalytic properties have been demonstrated to stem, in large part, from the efficient electron transfer at its edges rather than its basal planes. In fact, it has been reported that graphene exhibits in certain systems electrocatalytic activity similar to that of edge plane highly ordered pyrolytic graphite. In addition to having a highly expanded network, an interconnected corrugated carbon-based network also displays a large amount of edge planes (Refer back to FIG. 5), making it an ideal system for studying the role of edge planes on the electrochemistry of graphene-based nanomaterials. In this regard, the electrochemical behavior associated with the electron transfer of flexible electrodes made of interconnected corrugated carbon-based networks using a [Fe(CN)6]3−/4− couple as a redox probe is characterized. For example, FIG. 15 compares the CV profiles of GO, graphite and electrodes made of interconnected corrugated carbon-based networks in an equimolar mixture of 5 mM K3[Fe(CN)6]/K4[Fe(CN)6] dissolved in 1.0 M KCl solution at a scan rate of 50 mV/s. Unlike GO and graphite, the electrode made of interconnected corrugated carbon-based networks approaches the behavior of a perfectly reversible system with a low ΔEp (peak-to-peak potential separation) of 59.5 mV at a scan rate of 10 mV/s to 97.6 mV at a scan rate 400 mV/s. The low ΔEp values approaches the calculated theoretical value of 59 mV. Given that ΔEp is directly related to the electron transfer rate constant (k0obs), the low experimental value of ΔEp indicates a very fast electron transfer rate. The calculated k0obs values vary from 1.266×104 cm s−1 for graphite and, as expected, increases for an interconnected corrugated carbon-based network to 1.333×10−2 cm s−1. The redox system that was used for the evaluation of the electron transfer kinetics was 5 mM K3[Fe(CN)6]/K4[Fe(CN)6] (1:1 molar ratio) dissolved in 1.0 M KCl solution. To ensure a stable electrochemical response, the electrodes were first cycled for at least 5 scans before collecting the experimental data. The heterogeneous electron transfer rate constant (k0obs) was determined using a method developed by Nicholson, which relates the peak separation (ΔEp) to a dimensionless kinetic parameter ψ, and consequently to k0obs according to the following equation: k o ⁢ b ⁢ s 0 = ψ ⁡ [ D O ⁢ π ⁢ ⁢ v ⁡ ( nF R ⁢ T ) ] ⁢ ( D R D O ) α 2 where DO and DR are the diffusion coefficients of the oxidized and reduced species, respectively. The other variables include v—the applied scan rate, n—the number of electrons transferred in the reaction, F—the Faraday constant, R—the gas constant, T—the absolute temperature and α—the transfer coefficient. The diffusion coefficients of the oxidized and reduced species are typically similar; therefore, the term (DR/DO)α/2 is ˜1. A diffusion coefficient (DO) of 7.26×10−6 cm2 s−1 was used for [[Fe(CN)6]3−/4− in 1.0 M KCl. In addition to the relatively large increase in the electron transfer rate at the electrode made of interconnected corrugated carbon-based networks (˜two orders of magnitude times faster than a graphite electrode), there is also substantial electrochemical activity for the electrode made of interconnected corrugated carbon-based networks as seen by an increase of ˜268% in the voltammetric peak current. These drastic improvements are attributed to the expanded architecture of interconnected corrugated carbon-based network films, which provide large open areas for the effective diffusion of the electroactive species and allow a better interfacial interaction with the interconnected corrugated carbon-based network surface. Additionally, it is surmised that the amount of edge-like surface per unit mass is thus, much higher than graphite, and therefore contributes to the higher electron transfer rates, as seen here. Given the large number of exposed edge sites in interconnected corrugated carbon-based networks, it is not surprising to find that it not only has a higher k0obs value than graphite, but surpasses that of carbon nanotube based electrodes and that of stacked graphene nanofibers. Note that the electrodes made of interconnected corrugated carbon-based networks are fabricated on flexible PET substrates covered with GO which, when laser reduced, serves as both the electrode and the current collector, thus making this particular electrode not only lightweight and flexible, but also inexpensive. In addition, the low oxygen content in interconnected corrugated carbon-based networks (˜3.5%) as shown through XPS analysis is quite advantageous to the electrochemical activity seen here, since a higher oxygen content at the edge plane sites have been shown to limit and slow down the electron transfer of the ferri-/ferrocyanide redox couple. As such, embodiments of the present disclosure provides methodologies for making highly electroactive electrodes for potential applications in vapor sensing, biosensing, electrocatalysis and energy storage. The present disclosure relates to a facile, solid-state and environmentally safe method for generating, patterning, and electronic tuning of graphite-based materials at a low cost. Interconnected corrugated carbon-based networks are shown to be successfully produced and selectively patterned from the direct laser irradiation of GO films under ambient conditions. Circuits and complex designs are directly patterned on various flexible substrates without masks, templates, post-processing, transferring techniques, or metal catalysts. In addition, by varying the laser intensity and laser irradiation treatments the electrical properties of interconnected corrugated carbon-based networks are precisely tuned over five orders of magnitude, a feature that has proven difficult with other methods. This new mode of generating interconnected corrugated carbon-based networks provides a new venue for manufacturing all organic based devices such as gas sensors, and other electronics. The relatively inexpensive method for generating interconnected corrugated carbon-based networks on thin flexible organic substrates makes it a relatively ideal heterogeneous scaffold for the selective growth of metal nanoparticles. Moreover, the selective growth of metal nanoparticles has the potential in electrocatalyzing methanol fuel cells. Further still, films made of interconnected corrugated carbon-based networks show exceptional electrochemical activity that surpasses other carbon-based electrodes in the electron charge transfer of ferri-/ferrocyanide redox couple. The simultaneous reduction and patterning of GO through the use of an inexpensive laser is a new technique, which offers significant versatility for the fabrication of electronic devices, all organic devices, asymmetric films, microfluidic devices, integrated dielectric layers, batteries, gas sensor, and electronic circuitry. In contrast to other lithography techniques, this process uses a low-cost infrared laser in an unmodified, commercially available CD/DVD optical disc drive with LightScribe technology to pattern complex images on GO and has the additional benefit to simultaneously produce the laser converted corrugated carbon network. A LightScribe technology laser is typically operated with a 780 nm wavelength at a power output within a range of around 5 mW to around 350 mW. However, it is to be under stood that as long as the carbon-based oxide absorbs within the spectrum of the laser's emission, the process is achievable at any wavelength at a given power output. This method is a simple, single step, low cost, and maskless solid-state approach to generating interconnected corrugated carbon-based networks that can be carried out without the necessity of any post-processing treatment on a variety of thin films. Unlike other reduction methods for generating graphite-based materials, this method is a non-chemical route and a relatively simple and environmentally safe process, which is not limited by chemical reducing agents. The technique described herein is inexpensive, does not require bulky equipment, displays direct control over film conductivity and image patterning, can be used as a single step for fabricating flexible electronic devices, all without the necessity for sophisticated alignment or producing expensive masks. Also, due to the conductive nature of the materials used, it is possible to control the resulting conductivity by simply patterning at different laser intensities and power, a property that has yet to been shown by other methods. Working circuit boards, electrodes, capacitors, and/or conducting wires are precisely patterned via a computerized program. The technique allows control over a variety of parameters, and therefore provides a venue for simplifying device fabrication and has the potential to be scaled, unlike other techniques that are limited by cost or equipment. This method is applicable to any photothermically active material, which includes but is not limited to GO, conducting polymers, and other photothermically active compounds such as carbon nanotubes. As described above, a method has been presented for producing graphite-based materials that is not only facile, inexpensive and versatile, but is a one step environmentally safe process for reducing and patterning graphite films in the solid state. A simple low energy, inexpensive infrared laser is used as a powerful tool for the effective reduction, subsequent expansion and exfoliation and fine patterning of GO. Aside from the ability to directly pattern and effectively produce large areas of highly reduced laser converted graphite films, this method is applicable to a variety of other thin substrates and has the potential to simplify the manufacturing process of devices made entirely from organic materials. A flexible all organic gas sensor has been fabricated directly by laser patterning of GO deposited on thin flexible PET. An interconnected corrugated carbon-based network is also shown to be an effective scaffold for the successful growth and size control of Pt nanoparticles via a simple electrochemical process. Finally, a flexible electrode made of interconnected corrugated carbon-based networks was fabricated, which displays a textbook-like reversibility with an impressive increase of ˜238% in electrochemical activity when compared to graphite towards the electron transfer between the ferri-/ferrocyanide redox couple. This proof-of concept process has the potential to effectively improve applications that would benefit from the high electrochemical activity demonstrated here including batteries, sensors and electrocatalysis. Those skilled in the art will recognize improvements and modifications to the embodiments of the present disclosure. All such improvements and modifications are considered within the scope of the concepts disclosed herein and the claims that follow.
060524243	claims	1. A piercing welding method for fabricating a double-wall structure having an inner wall, an outer wall surrounding said inner wall, and reinforcing ribs that connect said outer wall and said inner wall, the method comprising externally applying an electron beam at a right angle to the outer wall such that the applied electron beam penetrates the outer wall to reach an abutting rib, said abutting rib having a rib width, whereupon the outer wall is welded to the rib to form a welded structure having at least two piercing weld beads that are spaced apart by an unwelded area, wherein each of said at least two weld beads has a root and a weld bead width, wherein the sum of the widths of the at least two weld beads is at least 25% of the rib width, and wherein a length of the unwelded area exterior to the root of each bead is no more than 20% of the rib width. 2. The method according to claim 1, wherein the double-wall structure is the hull of ship, the fuselage of an airplane, a car body, a bridge structure, one of various containers in industrial plants, or a vacuum vessel. 3. The method according to claim 2, wherein the vacuum vessel is one for use in a nuclear fusion device. 4. The method according to claim 1, wherein shields are inserted between the outer and inner walls of the double-wall structure. 5. The method according to any one of claims 1-4, wherein wraparound welding is applied to the piercing welded end of each rib. 6. The method according to claim 1 further comprising welding the inner wall to the abutting rib by a welding technique selected from the group consisting of arc welding, electron beam welding, and laser beam welding.
	claims	1. A masking assembly for partially masking a specimen that is subject to radiation imaging, comprising:sections of radiation blocking material connected to define a specimen holding frame and a single imaging window within the specimen holding, frame exposing the specimen to the radiation;wherein the sections of radiation blocking material further define an interior region that is so dimensioned to receive exterior contours of the specimen against internal faces of the holding frame,wherein the internal faces of the holding frame are separate from other faces of the holding frame defining the imaging window, andwherein at least a portion of the specimen is visible through the imaging window. 2. A masking assembly according to claim 1, wherein the internal faces of the sections track at least one profile of the specimen, the faces defining at least one void between at least one surface of the specimen and the masking assembly. 3. A masking assembly according to claim 2, wherein the internal faces connect to the exterior contours of the specimen to define a gap that is within a tolerance measurement. 4. A masking assembly according to claim 2, wherein the internal faces connect at an angle configured to be open toward the interior region and so dimensioned to receive the specimen within the angle. 5. A masking assembly according to claim 1, wherein the internal faces are configured to extend over a portion of the exterior contours of the specimen. 6. A system for imaging a specimen with radiation, the system comprising:a source of radiation to be directed to the specimen;a filter between the source of radiation and the specimen to attenuate the radiation to a preferred power and wavelength;a masking assembly for partially masking the specimen, the masking assembly comprising:sections of radiation blocking material connected to define a specimen holding frame and an imaging window within the specimen holding frame,wherein the sections of radiation blocking material further define an interior region that is so dimensioned to receive exterior contours of the specimen against internal faces of the holding frame,wherein the internal faces of the holding frame are separate from other faces of the holding frame defining the imaging window, andwherein at least a portion of the specimen is visible through the imaging window. 7. A system according to claim 6, further comprising a digital imaging device comprising an imaging surface configured to receive the specimen surrounded by the masking assembly thereon,wherein the internal faces of the sections track at least one profile of the specimen, the faces enclosing at least one void between at least one surface of the specimen, the masking assembly, and the digital imaging device. 8. A system according to claim 7, wherein the digital imaging device further comprises a direct digital array receiving the radiation from the source, wherein the radiation traverses the imaging window of the masking assembly, the specimen, and is exposed to an imaging surface of the direct digital array. 9. A system according to claim 7, wherein the filter, the masking assembly, and the digital imaging device are configured to utilize the lowest power level of the radiation source to achieve an image of the specimen. 10. A method of masking a specimen for radiation imaging, comprising:attaching sections of radiation blocking material to one another such that the connected sections define a specimen holding frame and an imaging window within the specimen holding frame,arranging the sections of radiation blocking material to further define an interior region that is so dimensioned to receive exterior contours of a specimen against internal faces of the holding frame,wherein the internal faces of the holding frame are separate from other faces of the holding frame defining the imaging window, andwherein at least a portion of the specimen is visible through the imaging window. 11. A method according to claim 10, wherein the internal faces of the sections track at least one profile of the specimen, the faces defining at least one void between at least one surface of the specimen and the masking assembly. 12. A method according to claim 10, wherein the internal faces connect to the exterior contours of the specimen to define a gap that is within a tolerance measurement. 13. A method according to claim 10, wherein the internal faces define at least one angle that overlaps the specimen. 14. A method according to claim 10, further comprising dimensioning the masking assembly such that upon placing the specimen and the masking assembly on an imaging surface, the arrangement of the masking assembly, the imaging surface, and the exterior surfaces of the specimen enclose at least one void defined by the arrangement. 15. The system according to claim 6, further comprising additional sections of radiation blocking material that eliminate the imaging window, wherein the specimen is surrounded by radiation blocking material. 16. The system according to claim 15, wherein the holding frame blocks particular portions of radiation and is permeable to desired portions of radiation necessary for imaging.
	abstract	The present invention relates to a longitudinally divided emergency core cooling (ECC) duct in order to efficiently inject safety water to core of a pressurized light-water nuclear reactor. The ECC duct includes side supports for preventing the flow-induced vibration in the annular downcomer, and has structural stability while thermally expanding and contracting. A longitudinally divided ECC duct for emergency core cooling water injection of a nuclear reactor is provided on the periphery of a core barrel of a nuclear reactor, includes an emergency core cooling water inlet facing a direct vessel injection nozzle, and extends in a longitudinal direction of the core barrel. The longitudinally divided ECC duct is divided into a plurality of longitudinally-divided ducts in the longitudinal direction of the longitudinally divided ECC duct.
052241440	claims	1. A method of creating image information corresponding to an object which is inspected with penetrating radiation comprising, scanning said object with a flying spot of penetrating radiant energy along arcuate lines, detecting said radiant energy after interaction with said object, and as a result of said detecting, forming a detection signal, producing pixels from said detection signal, wherein each pixel corresponds to an elementary frontal area of the object, and feeding said pixels, or signals derived therefrom, to a utilization means in such manner that the pixels which correspond to a scan line of said object define an arcuate line in the utilization means. means for scanning said object with a flying spot of penetrating radiant energy along arcuate lines, radiant energy detecting means for receiving said radiant energy after interaction with said object and providing at least one detection signal, means for producing pixels from at least one detection signal, wherein each pixel corresponds to an elementary frontal area of the object, and means for feeding said pixels, or signals derived therefrom, to a utilization means. a source of penetrating radiation, an absorber plate having a fixed, arcuate slit therein, a chopper wheel made in part of radiation absorbing material, the chopper wheel having at least one radially oriented slit therein, the source of penetrating radiation, the absorber plate, and the chopper wheel being positioned with respect to each other so that radiation from the source passes through the fixed slit in the absorber plate and is incident upon the chopper wheel, and means for rotating the chopper wheel. scanning said object with a flying spot of penetrating radiant energy along curved scanning lines, detecting said radiant energy after interaction with said object, and as a result of said detecting, forming a detection signal, producing pixels from said detection signal, wherein each pixel corresponds to an elementary frontal area of the object, and feeding said pixels, or signals derived therefrom, to a utilization means in such manner that the pixels which correspond to a scan line of said object define a curved line in said utilization means which has the same shape as said curved scanning lines. 2. The method of claim 1 wherein said utilization means comprises a memory. 3. The method of claim wherein said utilization means comprises a display device. 4. An apparatus for creating image information corresponding to an object which is inspected with penetrating radiant energy, comprising, 5. The apparatus of claim 4 wherein said means for feeding said pixels or signals derived therefrom to a utilization means comprises means for feeding said utilization means in such manner that the pixels which correspond to a scan line of said object define an arcuate line in said utilization means. 6. The apparatus of claim 5 wherein said radiant energy detecting means comprises a transmission detector having a shape which corresponds to said arcuate lines. 7. The apparatus of claim 5 wherein said radiant energy detecting means comprises a detector of scattered radiation. 8. The apparatus of claim 5 wherein said means for scanning said object with a flying spot of radiant energy comprises a source of penetrating radiation, an absorber plate having a fixed slit in the shape of an arc, and an absorbing chopper wheel having at least one radially oriented slit. 9. The apparatus of claim 8 wherein said means for scanning said object with a flying spot of radiant energy further comprises means for effecting relative translation movement between said and object and said fixed slit in said absorber plate. 10. The apparatus of claim 5 wherein said utilization means comprises a storage means. 11. The apparatus of claim 5 wherein said utilization means comprises a display means. 12. An apparatus for scanning a object with penetrating radiation, comprising, 13. The apparatus of claim 12 wherein the chopper wheel has an annular ring shaped portion of radiation absorbing material, and the radial slits are in such portion. 14. The apparatus of claim 13 wherein the radiation from the source which passes through the fixed slit in the absorber plate is projected substantially wholly upon the annular ring shaped absorbing portion of the chopper wheel. 15. The apparatus of claim 12, further comprising means for effecting relative translation motion between said object and said fixed slit in said absorber plate. 16. A method of creating image information corresponding to an object which is inspected with penetrating radiation comprising,
	description	The present invention relates to a method for designing a nuclear fuel assembly which is intended to be positioned in a nuclear reactor, the assembly comprising a plurality of guide tubes, and a control cluster which itself comprises a plurality of control rods and a support for control rods, the control rods and the guide tubes extending in parallel with a longitudinal direction, each control rod being received in a guide tube in order to form pairs comprising guide tubes/control rods, each guide tube comprising a lower damping portion which comprises at least a portion of reduced inside diameter, which portion is intended to contain a fluid for damping the fall of the control rod which is received in the guide tube, the portion of reduced inside diameter surrounding the control rod with a radial passage gap when the control rod enters the guide tube. It will be appreciated that nuclear fuel assemblies must be dependable in order to allow reliable operation of nuclear reactors. Thus, design and construction provisions for such assemblies have been drawn up. These provisions impose a general framework and minimum criteria which the assembly constructors must take into consideration. As far as the guide tubes are concerned, the design provisions require verification that they remain stable under axial compression and that the mechanical integrity thereof is not affected during such compression. These criteria are aimed in particular at taking into consideration the axial compression of the guide tubes which results from contact against the upper bearing plate of the core by springs carried by an upper end piece of the assembly. Although the criteria imposed by the design provisions allow assemblies to be designed with satisfactory reliability, these are merely minimum criteria and it is desirable to develop a method which allows even more reliable assemblies to be designed. For all that, it is also desirable to reduce the safety margins during design in order to reduce the mass and the cost of the assemblies constructed. An objective of the invention is to overcome this problem by providing a method which allows nuclear fuel assemblies which are more reliable to be designed, while limiting the design margins. To this end, the invention relates to a method for designing a nuclear fuel assembly which is intended to be positioned in a nuclear reactor, the assembly comprising a plurality of guide tubes, and a control cluster which itself comprises a plurality of control rods and a support for control rods, the control rods and the guide tubes extending in parallel with a longitudinal direction, each control rod being received in a guide tube in order to form pairs comprising guide tubes/control rods, each guide tube comprising a lower damping portion which comprises at least a portion of reduced inside diameter, which portion is intended to contain a fluid for damping the fall of the control rod which is received in the guide tube, the portion of reduced inside diameter surrounding the control rod with a radial passage gap when the control rod is introduced in the guide tube, wherein the method comprises, for at least one pair comprising a guide tube/control rod, the following steps: a) establishing the falling speed of the control rod upon entry into the lower damping portion when the control cluster falls in the event of a shutdown of the nuclear reactor, b) establishing, based on the falling speed established in step a), the progression of the falling speed of the control rod in the lower damping portion, c) establishing, based on the progression of the speed established in step b), a maximum elevated pressure produced in the liquid contained in the lower damping portion, and d) establishing, based on the maximum elevated pressure established in step c), a maximum circumferential stress produced in the lower damping portion. According to specific embodiments, the method can comprise one or more of the following features, taken in isolation or according to all technically feasible combinations: the method further comprises a step for verifying, using the maximum circumferential stress established in step d), that a maximum stress admissible by the guide tube has not been exceeded, the establishing step b) is carried out using a higher value for the radial passage gap and the establishing step c) is performed using a lower value for the radial passage gap, the higher value is a maximum statistical value for the passage gap, the lower value is a minimum statistical value for the passage gap, the support of the control cluster comprising a helical spring for damping the impact of the support against an upper end piece of the assembly in the event of the control cluster falling during a shutdown of the nuclear reactor. The method further comprises the following steps:e) establishing the progression of the speed of the control cluster after the impact of the support against the upper end piece,f) establishing, based on the progression of the speed established in step e), a maximum longitudinal load for compression of the spring, andg) establishing, based on the maximum longitudinal load for compression, at least a maximum shearing stress in the spring, a maximum shearing stress is a shearing stress along the neutral axis of the spring, a maximum shearing stress is a shearing stress along the axis of the spring nearest the longitudinal center axis thereof, the method further comprises a step for verifying, using a maximum shearing stress established in step g), that a maximum stress admissible by the spring has not been exceeded. The invention further relates to a system for designing a nuclear fuel assembly, characterised in that it comprises an arrangement for performing the steps of a method as defined above. According to a variant, the system comprises a computer and storage arrangement, in which at least a program comprising instructions for performing steps of the method for designing a nuclear fuel assembly is stored. The invention further relates to a computer program comprising instructions for performing the steps of a method as defined above. The invention also relates to a medium which can be used in a computer and on which a program as defined above is recorded. FIG. 1 illustrates a nuclear fuel assembly 1 which mainly comprises a square-based lattice 2 for nuclear fuel rods 3 and a control cluster 4. The assembly 1 comprises grids 5 for maintaining the rods 3, which grids 5 are distributed over the height of the rods 3. A lower end piece 6 is arranged under the lower ends of the rods 3 and an upper end piece 7 above the upper ends of the rods 3. The upper end piece 7 is provided with springs 8 for pressing against the upper bearing plate of the reactor core, in which the assembly 1 is intended to be placed. The control cluster 4 comprises a plurality of control rods 10, for example, 24. Conventionally, the control rods 10 comprise a material which absorbs neutrons. The rods 3 and 10 extend in parallel with a vertical longitudinal direction L. The rods 10 are carried at the upper ends thereof by a support 11 which is generally referred to as a spider. As illustrated more particularly in FIG. 2, the spider 11 comprises a vertical central upper head 12 and a series of arms or vanes 13 which extend radially outwards from the lower end of the upper head 12 as far as the radially outer ends 14 thereof. Each control rod 10 is connected to an arm 13 at the upper end thereof. The upper head 12 of the spider 11 has a central blind hole 15 which opens towards the bottom and in which a damping helical spring 16 is received. The spring 16 extends vertically along a center axis A. A tightening screw 17 extends substantially over the entire height of the hole 15 and is screwed into the wall 18 delimiting the upper portion of the hole 15. The lower portion of the screw 17 extends through the base of a retaining ring 20 which rests on the lower end of the spring 16. The head 21 of the screw 17 rests, at the top, against the base of the retaining ring 20 in order to press the spring 16 against the wall 18 of the upper head 12. As illustrated in FIG. 3 for a control rod 10, each control rod 10 is received in a respective guide tube 24 which is arranged in the lattice 2 of fuel rods 3. In this manner, 24 pairs comprising a guide tube/control rod are formed. Since each of these pairs has a similar structure, only one will be described below. The guide tube 24 extends from the lower end piece 6 as far as the upper end piece 7. The guide tube 24 comprises a lower portion 26 of reduced inside diameter and an upper portion 27. The lower portion 26 is connected to the lower end piece 6 by a collared screw 28, through which a vertical through-hole 29 extends. The lower portion 26 of the guide tube 24 surrounds the control rod 10 with a radial passage gap J. The upper portion 27 is fixed to the upper end piece 7 and opens at the outside of the assembly 1. Lateral apertures 30, only one of which can be seen in FIG. 4, are provided in the upper portion 27 near the lower portion 26. When the assembly 1 is placed in a nuclear reactor, the cooling liquid of the reactor fills the interior of the guide tube 24. Conventionally, the control cluster 4 can be moved vertically relative to the remainder of the assembly 1 in order to allow adjustment of the reactivity during normal operation of the reactor, and therefore variations in power from zero power up to maximum output depending on the vertical introduction of the control rods 10 in the lattice 2 of rods 3. The vertical displacement of the control cluster 24 is conventionally performed by way of a drive rod which is connected to the upper end of the upper head 12. When the reactor is shut down, the drive rod and the assembly 4 fall due to gravity. At the start of this falling movement, the control rods 10 are guided only by the upper portions 27 of the guide tubes 24 and have not yet reached the lower portions 26. Once the falling action has ended, the lower ends of the control rods 10 are introduced in the lower portions 26. The cooling fluid contained in the portions 26 is then violently forced, on the one hand, upwards thereby and, on the other hand, downwards through the apertures 29 of the collared screws 28. Each lower portion 26 therefore behaves in the manner of a hydraulic damper braking the falling movement of the corresponding control rod 10, and therefore of the assembly 4. This braking phase ends at the end of the travel path with the impact of the spider 11 against the upper end piece 7 of the assembly 1. This impact is performed by a the retaining ring 20. During this impact, the spring 16 is compressed vertically in order to absorb the shock. According to the invention, the assembly 1 has been designed in order to take into consideration the specific stresses brought about in the assembly by the fall of the control cluster 4 during such a shutdown of the reactor. In this manner, in order to design the assembly 1, in particular a data-processing system 32 has been used, as illustrated schematically in FIG. 4. This system 32 comprises, for example, a computer or data processing unit 34 comprising one or more processors, storage arrangement 36, input/output arrangement 38, and optionally display arrangement 40. Instructions which can be performed by the computer 34 are stored in the form of one or more programs in the storage arrangement 36. These instructions are, for example, instructions in FORTRAN programming code. These various instructions, when they are performed by the computer 34, allow the method illustrated by the flow chart of FIG. 5 to be performed. In a first step illustrated by the box 42 of this figure, the computer 34 calculates, based on data 43, the progression of the falling speed of a control rod 10 in the upper portion 27 of the corresponding guide tube 24 in the event of a shutdown of the reactor. This calculation can be performed assuming, for example, that the control rod 10 is first subjected to constant loads: gravitational force: fg=Mg, Archimedes' thrust: fa=−p gV, pressure difference in the core: fc, and mechanical friction: fm,where M and V are the mass and the volume, respectively, of the assembly 4 and the drive rod thereof. The control rod 10 is also subjected to loads as a function of the speed or position thereof, for example, hydraulic friction which can be obtained from: fh=−cl (M+ρV) v2, with v=speed of the assembly 4 and therefore of the rod 10 in question. Thus, the equation of the movement of the rod in the upper portion 27 of the guide tube 24 is as follows: ( M + ρ ⁢ ⁢ V ) ⁢ ⅆ v ⅆ t = Σ ⁢ ⁢ f This gives: ⅆ v ⅆ t = c ⁢ ⁢ 2 - c ⁢ ⁢ 1 ⁢ v 2 with cl=hydraulic friction in the guide tube and c ⁢ ⁢ 2 = fg + fa + fc + fm M + ρ ⁢ ⁢ V C1 and c2 are, for example, experimental data measured during drop tests of the control cluster 4. These data are, with the other data necessary for the calculation, such as the mass and the volume of the assembly 4 and the drive rod thereof, introduced, for example, in the form of a file 43 by way of the input/output arrangement 38. The computer 34 resolves the equation of the movement of the control rod 10, for example, using the NEWTON method. Thus, the progression of the speed of the control rod 10 in the upper portion 27 is known as a function of time. The profile established in this manner can be displayed in the form of a curve by the display arrangement 40. This curve is illustrated by FIG. 6. In this manner, at the end of the step illustrated by the box 42, the speed of the control rod 10 is known at the point of entry to the lower damping portion 26 of the guide tube 24. Based on the results of the step of box 42, the computer 34 calculates the progression of the speed of the control rod 10 during its fall in the lower damping portion 26. This step is schematically illustrated by box 44. This step can be performed using the following equation: - ⅆ v ⅆ t = c ⁢ ⁢ 2 - ( cl + SCAxNCA ⁢ ⁢ Δ ⁢ ⁢ P M + ρ ⁢ ⁢ V ⁢ ⁢ v 2 ) ⁢ ⁢ v 2 ⁢ ⁢ with c ⁢ ⁢ 2 = fg + fa M + ρ ⁢ ⁢ V = M - ρ ⁢ ⁢ V M + ρ ⁢ ⁢ V ⁢ g SCA=cross-section of the rod 10 andNCA=number of rods 10 in the assembly 4. Therefore, the hypothesis that fc and fm, are negligible is applied here. The difference ΔP represents the elevated pressure produced in the cooling liquid contained in the guide tube 24, for example, the pressure thereof between the lower end of the rod 10 and the pressure present in the upper portion 27 of the guide tube 24. ΔP can be established by the following formula: Δ ⁢ ⁢ P = 1 2 ⁢ ρ ⁢ ⁢ Q 2 ⁢ v 2 ⁡ ( EXPA + CONTRA + FECRxCISAxz ) ⁢ ⁢ where EXPA = ( SCA SACM ⁢ ( 1 ⁢ ⁢ SACM SACTG ) ) 2 with SM=cross-section of the lower portion 26, SACM=SM−SCA=cross-section of the annular space between the rod 10 and the lower portion 26, SACTG=STG−SCA, where STG is the cross-section of the upper portion 27 of the guide tube 24, CONTRA = 0.4 ⁢ ( 1 ⁢ ⁢ SACM SM ) ⁢ ( SCA SACM ) 2 FECR=coefficient of loss of load owing to friction in the lower portion 26, CISA = ( SCA SM ) 2 ⁢ 1 DM DM=mean diameter of the guide tube 24 in the upper portion 27,z=height of the rod 10 introduced in the lower portion 26 of the guide tube 24, andQ=fraction of liquid flowing upwards out of the lower portion 26 The resolution of the equations governing the movement of the rod 10 after entry into the lower portion 26 is carried out by the computer 34, for example, using the RUNGE-KUTTA method. Thus, at the end of step 44, the progression of the speed of the control rod 10 in the lower portion 26 of the guide tube 24 is known before the impact of the spider 11 on the upper end piece 7. The speed profile established in this manner can be displayed, for example, by the arrangement 40, as illustrated in FIG. 7. On the curve in FIG. 7, the speed profile established during step 44 is the portion located to the left of the point 45. The computer 34 then performs, in the step of box 46, the calculation of the maximum elevated pressure produced ΔPMAX. This calculation can be performed, for example, based on the formula: Δ ⁢ ⁢ P = 1 2 ⁢ p ⁢ ⁢ Q 2 ⁢ v 2 ⁡ ( EXPA + CONTRA + FECRxCISAxz ) . The computer 34 performs, in the step 48, the calculation of a circumferential stress and maximum normal σθMAX, to which the lower portion 26 of the guide tube 24 is subjected due to the maximum elevated pressure ΔPMAX. This stress can be calculated based on the formula: σ θ ⁢ ⁢ MAX = 1 2 ⁢ Δ ⁢ ⁢ P MAX ⁡ ( DPM EMP + 1 ) where DPM=inside diameter of the lower portion 26 and EMP=minimum thickness of the wall of the lower portion 26. The system 32 can then provide, due to the input/output arrangement 38, a first result in the form of a file 49 containing the value σθMAX established, and optionally the maximum elevated pressure ΔPMAX established. Next, the system 32 performs the calculation of the progression of the speed of the control rod 10 after it comes into contact with the spider 11 and the upper end piece 7. This calculation step is illustrated by the box 50 in FIG. 5. This calculation can be performed, for example, using the following equation when the ring 20, and therefore the spider 11, is in contact with the upper end piece 7: ( M + ρ ⁢ ⁢ V ) ⁢ ⅆ v ⅆ t = ( M - ρ ⁢ ⁢ V ) ⁢ g - PRCH - K ⁡ ( z - LAI ) - c ⁢ ⁢ 3 ⁢ v with PRCH=preload of the spring 16=PRCMP×K, where PRCMP is the precompression of the spring 16 and K the rigidity of the spring 16,LAI=distance traveled by the control rod in the lower portion 26 before impact, andc3=coefficient of hydraulic damping in order to model the damping in the lower portion 26. In the event of a rebound, for example, when the spider 11 is no longer in contact with the upper end piece 7, the equation for movement of the control rod 10 in question is written as follows: ( M + ρ ⁢ ⁢ V ) ⁢ ⅆ v ⅆ t = ( M - ρ ⁢ ⁢ V ) ⁢ g - c ⁢ ⁢ 3 ⁢ v These two equations are integrated by the computer 34, for example, using the RUNGE-KUTTA method. Therefore, the step 50 allows the kinematics of the control cluster 4 to be established during the mechanical damping of the shock by the spring 16. The speed profile established in this manner can be displayed, for example, by the arrangement 40. This profile corresponds to the portion located to the right of the point 45 on the curve in FIG. 7. Based on the results of this step, the system 32 performs, in the step 52, the calculation of a maximum vertical compression force FMAX, to which the spring 16 is subjected during the mechanical damping. This calculation can be carried out, for example, based on the following formula:FMAX=MAX{K(z−LAI)+PRCH} The system 32 then performs, in the step of box 54, the calculation of an approximate maximum shearing stress τMAX in the spring 16: τ MAX = 8 ⁢ F MAX ⁢ DFN π ⁢ ⁢ DFR 3 with DFN=DER−DFR andDER=outside diameter of the spring 16,DFR=diameter of the wire of the spring 16. Subsequently, the system 32 can optionally perform, based on the maximum stress τMAX, the calculation of maximum corrected stresses. These stresses can be calculated by multiplying τMAX by different factors. Thus, it is possible to calculate: τ MAX ⁢ ⁢ 1 = τ MAX ⁢ xK c ⁢ ⁢ and τ MAX ⁢ ⁢ 2 = τ MAX ⁢ xK ⁢ ⁢ with Kc = 1 + 0.5 C , C = DFN DFR , and K = 4 ⁢ C - 1 4 ⁢ C - 4 + 0.615 C The stress τMAX1 corresponds to the shearing stress along the neutral axis FN (FIG. 2) of the spring 16. The stress τMAX2 corresponds to the stress along the axis F2 (FIG. 2) of the spring 16 nearest the vertical center axis A of the spring 16 (see FIG. 2). At the end of this step illustrated by the box 56, the system 32 provides the various maximum shearing stresses calculated, for example, in the form of data stored in a file 57, which are transmitted by the input/output arrangement 38. Based on the data contained in the files 49 and 57, which have also been stored in the storage means 36, the computer 34 will verify that the maximum stresses calculated are indeed acceptable for the materials which respectively constitute the guide tube 24 and the helical spring 16. This step has been schematically illustrated by the box 58 in FIG. 5. During such a step, the system 32 will, for example, verify that the maximum shearing stresses calculated during the steps 54 and 56 are less than maximum values admissible by the material which constitutes the spring 16. This verification is performed by a comparison of τMAX, τMAX1 and τMAX2 with a maximum value admissible by the material of the spring 16. As far as the maximum circumferential stress σθmax is concerned, the verification can be carried out based on a formula of the type:ƒ(σΘMAX)<σadmissible where σadmissible refers to the material which constitutes the lower portions 26 of the guide tubes 24. The function f can be a function which takes into consideration other stresses to which the guide tubes 24 can be subjected. Such a stress can be a vertical compression stress σA, to which the guide tubes 24 are subjected during the contact of the springs 8 of the upper end piece 7 against the upper bearing plate of the core in order to counterbalance the hydrostatic thrust during operation. Thus, the function f can be, for example, in the form ofƒ(σθMAX,σA)=σθMAX+σA It will be appreciated that this last step, illustrated by the box 58, can be performed by separate software which generally performs the validation of various design parameters of the assembly 1 based on results provided by various pieces of software each dedicated to taking into consideration specific operating conditions and which include the software which performs the steps 42, 44, 46, 48, 50, 52, 54 and 56. In general terms, the file 43 comprising the data 43 used by the method for the various calculations can comprise the data of Table 1 below. TABLE 1outside diameter of control rod 10(m)Nominal;maximuminside diameter of upper portion 27(m)Nominal;maximuminside diameter of lower portion 26(m)Nominal;maximumtotal length of lower portion 26(m)damping travel before impact(m)minimum thickness of wall of lower portion 26(m)maximum roughness of rod 10/tube 24(m)Nominal;tolerancediameter of aperture 29(m)length of aperture 29(m)roughness of aperture 29(m)moving mass IVI(kg)volumetric mass of liquid(kg/m3)kinematic viscosity of liquid(mz/s)c1(/m)c2(m/s2)Young's modulus of guide tube 24(Pa)Poisson's ratio of wide tube 24spring precompression 16(m)preloading of spring 16(N)length of spring 16 with contiguous turns(m)outside diameter of spring 16(m)diameter of wire of spring 16(m)compression when upper head 12 is in contact(m)with (upper end piece 7END Similarly, the file 49 comprising the results from step 48 can comprise the data of Table 2 below. TABLE 2ΔPMAX: maximum elevated pressure in lower portion 26(Pa)ZMAX: corresponding penetration in lower portion 26(m)σθMAX: maximum stress in lower portion 26(Pa)fmax: maximum force on lower end piece 6(N)tdur: duration of fall in lower portion 26 before impact(s)vfin: speed of impact of assembly 4 on upper end piece 7(m/s) The file 57 comprising the results of step 56 can itself contain the data of Table 3 below. TABLE 3FMAX: maximum compression force on spring 16(N)hMAX: maximum deflection of spring 16(m)τMAX: approximate maximum stress in spring(Pa)τMAX1: approximate maximum stress corrected by Kc(Pa)τMAX2: approximate maximum stress corrected by K(Pa)(Wahl coefficient) It has been possible to verify by experiment that the maximum elevated pressures and the maximum stresses obtained by means of steps 42, 44, 46 and 48 were reliable. In this manner, the first corresponding part of the method allows reliable guide tubes 24 to be designed. Furthermore, this first part calculates only a single stress which appears to be the pertinent stress for the conditions being considered. Consequently, this first part of the method allows the security margins to be limited during design, and therefore assemblies which are relatively light and economical to be designed. The second part of the method, which corresponds to steps 50, 52, 54 and 56, also allows maximum stresses to be reliably calculated, as confirmed by experiment. Thus, the second part of the method allows a reliable design to be arrived at by calculation for the spider springs 16, which design is found to be advantageous in comparison with the method of tests alone which is currently imposed by provisions. It will be appreciated that the second part of the method calculates only the small number of stresses, and in particular those on the axis F2 of the spring 16 nearest the center axis A of the spring, which are found to be pertinent to the conditions envisaged. In this manner, the second part of the method allows the design margins to be reduced. In more general terms, the steps 42, 44, 46 and 48, on the one hand, and 50, 52, 54 and 56, on the other, can be carried out by separate pieces of software. In order to increase the reliability of the calculation, for carrying out the first part of the method it is possible to use, as the passage gap J, the nominal value of the gap, or this nominal value corrected by the manufacturing tolerance, or a value resulting from statistical studies of the distribution of passage gaps J obtained in constructed assemblies. In a variant, it is possible to use a gap value J which is greater for steps 42 and 44 and a smaller gap value J for steps 46 and 48. This allows a high stress value σθMAX to be calculated because the speed reached during the fall of the rod 10 in question is high and the volume available in the lower portion 26 for the liquid during damping is small. However, this high stress value is not unrealistic and therefore does not lead to unjustified design margins, as illustrated by the following example. According to a specific variant, the upper value can be a maximum value for gap J which is verified with a given probability, for example, 95%, in constructed assemblies, and the lower value can be a minimum value obtained with the same probability. This variant allows an approximation of the situation where a single pair comprising a guide tube/control rod has minimum gap J, where the maximum stress σθMAX would be reached, and where all the other pairs comprising a guide tube/control rod have the maximum passage gap J, which would be the most extreme case. In some variants, the first part of the method could also take into consideration forms of the lower damping portion 26 which are different from those described previously. In this manner, these lower damping portions could have a plurality of successive portions of reduced diameter, optionally separated by portions of increased diameter, generally referred to as cavities. In some variants, the first part of the method is carried out with collared screws 28 which are not perforated by holes 29. In still more general terms, the first part and the second part of the design method described can be used independently of each other. In this manner, it is possible to carry out the second part relating to the design of the spring 16 without referring to the calculation of the elevated pressure ΔP and the stress σθMAX.
	abstract	A focused ion beam (FIB) system is disclosed, comprising an inductively coupled plasma ion source, an insulating plasma chamber containing the plasma, a conducting source biasing electrode in contact with the plasma and biased to a high voltage to control the ion beam energy at a sample, and a plurality of apertures. The plasma within the plasma chamber serves as a virtual source for an ion column comprising one or more lenses which form a focused ion beam on the surface of a sample to be imaged and/or FIB-processed. The plasma is initiated by a plasma igniter mounted near or at the column which induces a high voltage oscillatory pulse on the source biasing electrode. By mounting the plasma igniter near the column, capacitive effects of the cable connecting the source biasing electrode to the biasing power supply are minimized. Ion beam sputtering of the apertures is minimized by proper aperture materials selection.
	summary
052375943	claims	1. A method for characterizing at least one activated element in an earth formation surrounding a borehole, comprising the steps of: (a) displacing in said borehole a sonde comprising a neutron source and at least two gamma ray detectors longitudinally spaced from said source, while irradiating said formation with neutrons of sufficient energy to interact with said element according to the activation reaction; (b) detecting and counting at each detector the gamma rays resulting from the activation of atoms of said element; (c) determining, at each depth, the number of gamma ray counts detected during the time period defined by the time instants when respectively said source and said detectors pass that depth, said determination of gamma ray counts being made for each detector at each depth; (d) establishing a relationship, for each depth, between the counts from the respective detectors for that depth and the corresponding time instants when the corresponding detector passes that depth; and (e) deriving from said relationship at least one characteristic of said element. setting one detector as a reference detector; determining the time when another detector reaches a given depth by adding to the time when said reference detector reaches said depth, an additional time "Delta t" given by: EQU Delta t=d / V cross-plotting, for each depth, the gamma ray counts versus the corresponding times, one plot corresponding to one depth and including dots which forms a curve representative of said element at that depth; and deriving from said curve at least one characteristic of said element. (a) lowering a sonde comprising a neutron source and a gamma ray detector down to the bottom of said borehole or below the zone of earth formation to investigate; (b) pulling said sonde while irradiating said formation with neutrons of sufficient energy to interact with said element according to the activation reaction; (c) detecting and counting at each detector the gamma rays resulting from the activation of atoms of said element; (d) determining, at each depth, the number of gamma ray counts detected during the time period defined by the time instants when respectively said source and said detector pass that depth, said determination of gamma ray counts being made for each detector at each depth; (e) establishing a relationship, for each depth, between the counts from the respective detectors for that depth and the corresponding time instants when the corresponding detector reaches that depth; (f) repeating at least once the sequence including steps (a)-(e), said neutron source being turned off; and (g) deriving from said relationship at least one characteristic of said element. (1) neutron source means for irradiating said earth formation with neutrons of sufficient energy to interact with atoms of said element according to the activation reaction; (2) means for detecting and counting at least two locations longitudinally spaced from said source, the gamma rays resulting from the activation of atoms of said element; (3) means for determining, at each depth, the number of gamma ray counts detected during the time period defined by the time instants when respectively said source and said detector pass that depth, said determination being made for each detector at each depth; (4) means for establishing a relationship, for each depth, between the counts from the respective detectors for that depth and the corresponding time instants when the corresponding detector reaches that depth; and (5) means for deriving from said relationship at least one characteristic of said element. means for cross-plotting the gamma ray counts versus time, wherein said relationship shows approximately the form of a straight line; and means for deriving from the slope of said line the identity of said element. (a) means for lowering a sonde comprising a neutron source and a gamma ray detector down to the bottom of said borehole or below the zone of earth formation to investigate; (b) means for pulling said sonde while irradiating said formation with neutrons of sufficient energy to interact with said element according to the activation reaction; (c) means for determining, at each depth, the number of gamma ray counts detected at each detector during the time period defined by the time instants when respectively said source and said detector pass that depth, said determination being made of gamma ray counts for each detector at each depth; (d) means for establishing a relationship, or each depth, between the counts from the respective detectors for that depth and the corresponding time instants when the corresponding detector reaches that depth; (e) means for repeating at least once the sequence including steps a-e, said neutron source being turned off; and (f) means for deriving from said relationship at least one characteristic of said element. (a) irradiating said formation with neutrons of sufficient energy to activate atoms of at least one element in the formation; (b) detecting and counting the gamma rays resulting from the activation of atoms in said formation at at least two time instances; (c) determining from the counted gamma rays and time intervals the reduction rate of gamma rays; and (d) deriving from the gamma ray reduction ray the activated element of said formation. (a) displacing in said borehole a sonde comprising a neutron source and at least two gamma ray detectors longitudinally spaced from said source, while irradiating said formation with neutrons of sufficient energy to activate atoms of at least one element in the formation; (b) detecting and counting at each detector the gamma rays resulting from the activation of atoms in said formation; (c) determining the number of gamma ray counts detected during the time period defined by the time instants when respectively said source and each said detector pass through an activated region of said earth formation; (d) establishing a relationship between the counts from the respective detectors and the corresponding time instants when the detectors pass through the activated region of said formation; and (e) deriving from said relationship the activated element of said formation. cross-plotting, for each activation region, the gamma ray counts from said detectors versus the corresponding times, the plot including dots which form a curve representative of said activated element in said activated region; and deriving from said curve at least one characteristic of said element. 2. The method according to claim 1 further comprising: 3. The method according to claim 1 further comprising: 4. The method according to claim 3 wherein said curve is substantially a straight line the slope of which is representative of the identity of said element. 5. The method according to claim 1 wherein the counts for one detector are representative of the quantity of said element of the radial distance between said borehole and atoms of said element. 6. The method according to claim 1 wherein said sonde comprises four gamma ray detectors. 7. The method according to claim 1 wherein said detectors are disposed on the same side along the longitudinal axis of said sonde with respect to said neutron source. 8. A method for characterizing at least one activated element in an earth formation surrounding a borehole, comprising the steps of: 9. The method according to claim 1 wherein said element comprises aluminum, silicon, magnesium or gold. 10. A logging apparatus for characterizing an element of earth formation surrounding a borehole, comprising a sonde comprising: 11. The apparatus according to claim 10 further comprising means for determining one time instant for a first detector including means for adding to the time instant of a second detector, the time "t" needed for the latter to reach that depth given by: t=d / V, where "d" is the spacing between the first and the second detector and "V" is the speed at which said sonde is displaced in said borehole. 12. The apparatus according to claim 10 further comprising: 13. The apparatus according to claim 10 wherein the counts for one detector are representative of the quantity of said element and of the radial distance between said borehole and atoms of said element. 14. The apparatus according to claim 10 wherein said sonde comprises four gamma ray detecting means. 15. The apparatus according to claim 10 wherein said detecting means are disposed on the same side along the longitudinal axis of said sonde, with respect to said neutron source. 16. The apparatus according to claim 15 wherein said neutron source is disposed above said detecting means. 17. An apparatus for characterizing at least one activated element in an earth formation surrounding a borehole, comprising: 18. The method according to claim 1 wherein the relationship in step (d) is established by cross-plotting for each detector the gamma ray counts versus the corresponding instant of time for that count. 19. A method for characterizing an element of an earth formation surrounding a borehole by activating atoms in said element comprising the steps of: 20. The method according to claim 19 wherein the gamma ray reduction rate is determined by plotting each gamma ray count versus the corresponding instant of time for that count. 21. The method according to claim 20 wherein the activation element is derived by comparing the gamma counts versus time plot to predetermined plots of various elements to determined the closest element match. 22. A method of characterizing an activated element in an earth formation surrounding a borehole by activating atoms in said element comprising the steps of: 23. The method according to claim 22 wherein the relationship of step d is established by plotting for each detector, the gamma ray count for that detector versus the corresponding instant of time that count. 24. The method according to claim 22 wherein one detector is set as a reference detector and further comprising before step c the step of determining the time when another detector reaches the activated region, an additional time "Delta t" given by: EQU Delta t=d/V 25. The method according to claim 22 wherein the counts for one detector are representative of the quantity of said activated element and of the radial distance between said borehole and atoms of said activated element. 26. The method according to claim 40 further comprising: 27. The method according to claim 26 wherein said curve is substantially a straight line, the slope of which is representative of the identity of said activated element.
	summary
060581549	description	DESCRIPTION OF THE PREFERRED EMBODIMENTS As shown in FIG. 2, the invention appears particularly simple to construct and implement and the container assembly has a perfectly smooth, cylindrical outer surface and its entire length is free from rough projections. In FIG. 2, reference (1) is the thin wall of the tubular body of the multi-purpose canister, (2) is the inner cavity designed to house the irradiated fuel assemblies placed in an appropriate holder. The upper end (10) of the tubular body may have the same thickness as the body but here it has been welded on and has a slight additional thickness inside said tubular body. Like the tubular body it is circular in cross-section except for certain places, for example when the shoulder (11) does not run around the entire periphery of the inner wall of the tubular body, as will be seen below. The external walls of the tubular body (vertical in this example) and its upper end (10), which may be added on, are aligned so that there are no projections capable of causing an obstruction. The optional internal overthickness gives the upper end (10) of the unit increased rigidity so that the guaranteed leaktightness of the apparatus according to the invention is improved without reducing the capacity of said unit. The upper end (10) has a shoulder (11) cut into the inner wall; this shoulder may be cut into the entire periphery of said inner wall or only part thereof. The shoulder ensures that the support for the bearing disk or collar (12) is stable. The bearing disk or collar is circular and has an upper surface, a lower surface and a cylindrical lateral wall. On the upper surface of the circular bearing disk or collar (12) is placed a first solid disk (13). The periphery of said solid disk (13) essentially of circular shape comprises sealing means comprising a bevel (14) cut into the periphery of its lower edge and an O-ring (15) that is usually made of metal placed in the housing created by said bevel (14). The metal seal may be of the Helicoflex.TM. or any other type of metal seal having similar properties. In another embodiment the first solid disk (13), which is in the lower position, bears directly on the shoulder (11). Leaktightness is ensured by the O-ring (15) between the circular cross-section vertical cylindrical inner wall of the upper end (10), or the end of tubular body (1) where no specific end-piece has been added, and either the surface of the bevel or the upper horizontal surface of the support disk. This embodiment using a support disk is particularly advantageous since it compresses the O-ring (15) and ensures leaktightness when the shoulder (11) is not cut into the entire periphery of the inner wall. It achieves this in particular because said leaktight seal is partly obtained due to pressure of the seal on the vertical section of the inner wall. In order to both guarantee improved leaktightness and improve testing of the leaktightness obtained, a second solid disk (13') that is identical to first solid disk (13), complete with bevel (14') and metal seal (15') is placed on the upper surface of the disk (13). Seal (15') is also compressed due to the action of the upper surface of the first solid disk (13) in lower position on which it is placed. The gripping apparatus (16) is entirely located inside the upper end (10) of tubular body (1). This is normally of known type and comprises gripping means that operate in conjunction with the inner wall, thereby making the closure of the unit leaktight by compressing the seals (15, 15') and pressing vertically on the upper disk. It will be seen that, as the assembly is tightened, the seals (15, 15') are compressed to create a leaktight seal between the bevels of the first and second solid disks (13) and (13') and the inner wall of the upper end (10) of the end of the tubular body (1). The number of solid disks may simply be increased to give further improvements in the degree of leaktightness required. It will be seen that the reversible closure apparatus of the invention is easy to produce. It also allows a multi-purpose canister to pass all the way through an over-pack from one end to the other and also gives a high degree of leaktightness and mechanical resistance that meets the requirements of the safety regulations.
	abstract	A beam transport system for a hadron therapy facility comprises: a main beam transport line; secondary beam transport lines branching off from the main beam transport line for delivering the hadron beam into the patient treatment stations; and switching electromagnets for deviating the hadron beam from the main beam transport line into the secondary beam transport line. A discharge circuit associated with each switching electromagnet comprises a discharge accelerating circuit capable of generating a voltage opposing the counter electromotive force induced in the electromagnet coil of the switching electromagnet when the energization of the electromagnet coil producing the hadron beam deviation is interrupted, wherein this voltage stays substantially constant or increases as the current induced in the electromagnet coil decreases.
	summary
	description	This application claims priority under 35 U.S.C. §119(a) to a French Patent Application filed Dec. 12, 2011, and assigned Serial No. FR.11/61500, the entire disclosure of which is incorporated herein by reference. 1. Field of the Invention The present invention concerns a method and device for processing waste issuing from the nuclear industry. It relates more particularly to a method and device for limiting the degassing of tritiated waste produced by nuclear reactions. 2. Description of the Related Art Tritium (T or 3H) may be present in several types of nuclear reactor. In the majority of fission reactors, it may for example be produced by the ternary fission of uranium (U) and plutonium (Pu) as well as in neutron reactions between boron (B) and lithium (Li). In addition, in fusion reactors, tritium is used as a fuel. That is the case of Joint European Torus (JET) reactors, for International Thermonuclear Experimental Reactors (ITER) and (Demonstration Power Plant (DEMO reactors) designed to succeed the ITER reactor. In these reactors, the fusion reaction from deuterium (D or 2H) and tritium (T or 3H) is governed by Equation (1):D+T→42He+01n+17.6 MeV (1) One of the properties of tritium, just like the other isotopes of hydrogen, is its ability to penetrate through materials, which contaminates them. This property makes tritiated waste very specific since both the radioactive elements and the degassing must be taken into account in the management of the waste. Moreover, tritium may be present in two gaseous forms, tritiated hydrogen (HT or T2) and tritiated water (HTO or T2O). It should be noted that tritiated hydrogen (HT or T2) is much more mobile than tritiated water (HTO or T2O), even in vapour form. Tritiated hydrogen (HT or T2) also has a very small size. Its great mobility and its small size promote its diffusion even through the finest porosities, passing for example through rubber and diffusing in the majority of types of steel and concrete. Trapping tritiated hydrogen (HT or T2) is therefore made particularly complex. With the development of fusion reactors, the quantity of tritiated waste will thus increase significantly. There consequently exists a particularly great need consisting of sizing an effective method of detritiation and limitation of degassing, in particular at ambient temperature and atmospheric pressure. The reduction in the quantity of tritiated hydrogen produced by a package of waste must be sufficiently effective to satisfy the existing acceptability criteria of warehousing and storage centres, so that these packages comprising tritiated waste can be included in the existing warehousing and storage centres. One solution has been proposed and is described in U.S. Pat. No. 5,464,988. This patent proposes a device for storing and transporting tritiated water contained in a gas. The device comprises an external drum and an internal drum placed inside the external drum. Inside the internal drum is a disposable molecular sieve intended to be loaded with tritium. The device also comprises entry and exit diffusers as well as a catalytic hydrogen recombiner. A protective and thermal insulation material is arranged in each space available between the two drums. The catalytic hydrogen recombiner converts the tritiated hydrogen into tritiated water, which can then be trapped by the molecular sieve in solid form with a view to storage and transportation thereof. This device affords a solution for trapping tritiated water in gaseous form. Trapping tritiated hydrogen (HT or T2) does however prove to be more complicated because of the smaller size and greater mobility of this gas compared with tritiated water in gaseous form. In addition, this solution is based on a particularly complex drum requiring numerous specific items of equipment that require specific maintenance. For example, the use of a catalytic recombiner requires a heating device and therefore maintenance of the system. In addition, human intervention is necessary for regularly changing the molecular sieve saturated with tritiated water. The complexity of this solution tends to make its cost relatively high and its reliability limited. There therefore exists a need consisting of proposing a simple and effective solution for limitation of the degassing of tritiated hydrogen (T2 or HT) and tritiated water (HTO or T2O), this solution also having to make it possible to reduce the maintenance and human intervention requirements. The solution must also be effective over several hundreds of years. The objective of the present invention is to describe such a solution. In the remainder of the disclosure of the invention tritiated water will be spoken of indifferently, which designates the elements HTO and T2O. In the present description, tritiated water designates tritiated water in vapour or liquid form. The present invention provides a method for reducing the quantity of tritiated hydrogen (T2 or HT) and/or tritiated water (HTO or T2O) generated by at least one package comprising at least one piece of waste. The method includes placing the package in contact with a mixture comprising manganese dioxide (MnO2) combined with a component comprising silver, and then a step of placing the package in contact with a molecular sieve. The mixture can effectively oxidize the tritiated hydrogen (T2 or HT), in order to reduce the diffusivity of this gas. The result is tritiated water (T2O or HTO), tritiated water being much less mobile than tritiated hydrogen. The tritiated water is then trapped by the molecular sieve. It has proved that the combination of the molecular sieve with the mixture comprising manganese dioxide associated with silver reduces the degassing of tritiated waste very effectively and safely even for small quantities of tritiated hydrogen. In particular, the combination of the method according to the invention provides high kinetics and great reactivity compared to tritiated hydrogen. In addition, this method of trapping tritiated hydrogen does not require any external maintenance. Particularly advantageously, it may be carried out at ambient pressure and temperature. Moreover, the reaction generates a thermally stable product. It is also reversible with difficulty or not at all over at least several hundreds of years. Advantageously, the trapping method proposed by the invention is relatively simple to implement and inexpensive. The present invention is therefore particularly relevant for use on an industrial scale. In the context of the present invention, tritiated waste means any radioactive waste liable to contain or degas tritiated hydrogen (HT or T2). The waste may also contain tritium in other forms such as for example tritiated water (HT or T2O). Typically, the waste issues from the nuclear industry and is liable to contain or degas tritiated hydrogen and/or tritiated water. In the context of the present invention, putting at least two elements, typically tritiated hydrogen or tritiated water, in contact with the mixture comprising manganese dioxide (MnO2) combined with a component comprising silver and/or with the molecular sieve means that these elements are placed so as to be able to react with one another. This putting in contact may therefore be a physical contact of two elements in the solid, liquid or gas state. In any event, a putting in contact enables the elements to interact so that the reaction of transformation of the tritiated hydrogen into tritiated water and the reaction of trapping the tritiated water by the molecular sieve take place. Optionally, the method according to the invention can have at least any one of the optional steps and features stated below. The present invention provides a package including at least one piece of waste generated by a nuclear fission reaction. Alternatively or cumulatively, the package may comprise at least one piece of waste formed by a spent fuel or one to be used for a nuclear fusion reaction. The package may also comprise a piece of waste made tritiated by the presence of another piece of tritiated waste or by the presence of a tritiated fuel. The waste may also be the product or fuel of a nuclear reaction. It may also be any element contaminated by the product or fuel of a nuclear reaction. The waste may thus be an object made radioactive such as a garment or a tool, reactor parts, radioactive fuel, etc. The component comprising silver comprises at least one of the following components: AgO, Ag2O, silver salts of the AgCl or AgNO3 type, or complexes comprising silver. Preferentially, the silver is in silver oxide form (Ago or Ag2O) in the mixture. The mass concentration of manganese dioxide in the mixture is ranging from 80% to 99% and the mass concentration of silver oxide Ag2O in the mixture is ranging from 20% to 1%. This corresponds to a mass fraction of silver (Ag) of between 0.93% and 18.6%. This is because 10% Ag2O corresponds to 9.3% silver (Ag). More precisely, the mass concentration of manganese dioxide in the mixture is ranging from 87% to 93% and the mass concentration of silver oxide Ag2O in the mixture is ranging from 13% to 7%. Even more preferentially, the mass concentrations in the mixture for manganese dioxide and silver oxide Ag2O are around 90% and 10%, respectively. Preferably, the silver is in the silver nitrate form (AgNO3) and the mass fraction of silver (Ag) in the mixture is ranging from 1.5% to 30% and preferably around 15%. According to a particular embodiment, the mixture comprises platinum (Pt) or a compound containing platinum. For reasons of clarity, in the remainder of the present description this compound containing platinum will be referred to as a “platinum compound”. Advantageously, it is platinum black 10% Pt. As is known, platinum black 10% Pt is composed of 90% active carbon and 10% platinum. Platinum accelerates and facilitates the oxidation of tritiated hydrogen. Advantageously, the mass concentration of the platinum compound in the mixture is ranging from 0.1% to 1%, that is to say a proportion of platinum ranging from 0.01% to 0.1% for platinum black 10% Pt. Preferably, this concentration is 0.5% Pt black 10%, that is to say 0.05% platinum, the manganese dioxide concentration is 89.3% and the silver oxide Ag2O concentration is 10.2%. More precisely, these concentrations are 0.56%, 89.28% and 10.16%, respectively. According to a particular embodiment, the mixture comprises solely manganese dioxide and silver oxide. According to another embodiment, the mixture comprises solely manganese dioxide, silver oxide and platinum compound. Preferentially, the molecular sieve is a 4A-type or 5A-type zeolite. Preferably the method comprises, prior to the steps of putting the package in contact with the mixture and putting the package in contact with at least one molecular sieve, a step of depositing the mixture on the molecular sieve. According to an advantageous embodiment, the step of depositing the mixture on the molecular sieve is accompanied by a step of mechanical fixing of the mixture on the molecular sieve. According to another advantageous embodiment, the step of depositing the mixture on the molecular sieve is accompanied by a step of chemical fixing of the mixture on the molecular sieve by means of a binder, such as for example water. In this case, the fixing step consists of putting the mixture and the molecular sieve in contact in aqueous phase followed by a step of drying at a temperature ranging from 150° to 200° C. for a period of between 12 h and 48 h. According to an embodiment in which the mixture is fixed to the molecular sieve, the package is in the form of a drum having a bottom and able to contain at least one piece of tritiated waste, the method comprising a step of placing, in the bottom of the drum, the sieve to which the mixture is fixed. According to an embodiment in which the mixture is not fixed to the molecular sieve, the package is in the form of a drum having a bottom and containing at least one piece of tritiated waste, the method comprising a step of placing the sieve in the bottom of the drum and a step of placing the mixture in the waste or on the waste. The sieve and the mixture can also be placed in a drum already completely or partially filled. According to yet another embodiment the mixture is placed on a metal mesh to form a covering, the method also comprising a step of enveloping at least one part of the package with the covering. Preferably, the metal mesh is flexible to enable the cover to match the shapes of the package. Preferably, the cover incorporates the molecular sieve, the latter being covered by the mixture. This embodiment is particularly advantageous when the package is in the form of waste coated in a matrix, such as cement, glass or bitumen matrix. The method comprises, prior to the step of putting the package in contact with a mixture of manganese dioxide (MnO2) combined with silver (Ag), a step of preparing the mixture during which the manganese dioxide (MnO2) is combined with a compound comprising silver such as for example a silver oxide or a silver salt, these examples not being limitative. Preferentially, during the step of preparing the mixture, the silver is added in the form of silver oxide (AgO or Ag2O) to the manganese dioxide (MnO2). According to a first embodiment, the step of preparing the mixture comprises a step of mixing a manganese dioxide powder with a silver oxide powder (AgO or Ag2O). Optionally, the method comprises, during the step of preparing the mixture and after the step of mixing the manganese dioxide powder with the silver oxide powder, a step of adding water to the mixed powders. The addition of water has the advantage of facilitating the spreading of the silver on the surface of the manganese dioxide. According to a second embodiment, the step of preparing the mixture comprises a step spreading a saline solution comprising silver ions (Ag+) on manganese dioxide in the solid state. Preferably, the manganese dioxide is in the form of a dusty powder. According to a third embodiment, the step of preparing the mixture comprises a step of immersing manganese dioxide in the solid state, preferably in a dusty powder form, in a solution comprising a salt or a silver oxide. More precisely, the silver is in the form of a cation Ag+ or Ag2+. One advantage of this method is that it is particularly effective for obtaining a homogeneous distribution of the silver on the manganese dioxide. According to a fourth embodiment, the step of preparing the mixture comprises a step of depositing silver on the manganese dioxide by a precipitation reaction of a solution comprising silver. The advantage of a method of impregnation by a liquid is the homogeneity of the final mixture obtained. The solution comprising silver is a solution of silver salts (AgCl, AgBr, AgNO3, etc.). The precipitation is caused by the introduction of sodium hydroxide (NaOH) in this silver nitrate solution. According to one embodiment, the package comprises a drum comprising a plurality of pieces of tritiated waste. Alternatively, the package comprises a piece of waste coated in a matrix, the matrix being for example concrete, bitumen or glass. Another subject matter of the present invention is a device for reducing the quantity of tritiated hydrogen, the device comprising at least one molecular sieve, with the device including a mixture comprising manganese dioxide (MnO2) combined with a compound comprising silver such as for example a silver oxide or a silver salt, these examples not being limitative. For the reasons mentioned previously in relation to the method, the invention thus proposes a device that is particularly effective for reducing the degassing of tritiated waste with a view to warehousing and storage thereof. According to an advantageous embodiment, the mixture comprising manganese dioxide (MnO2) combined with a compound comprising silver is fixed to the molecular sieve. Optionally, the layer has a thickness of between 5 and 20 μm. One advantage of this embodiment is enabling a homogeneous spreading of the various constituents, which improves the efficacy of the device. Preferably, the molecular sieve is a 4A-type or 5A-type zeolite. Advantageously, the layer is fixed to the sieve mechanically and/or by means of an additive. This additive is preferably a binder, such as for example water. According to another advantageous embodiment, the device comprises at least one drum able to contain at least one piece of tritiated waste as well as the sieve and the mixture. Optionally, the device includes at least one of: a drum containing waste with the mixture spread on the waste or in the middle of the waste; a sieve placed in the bottom of the drum; and the mixture at least partly covers the molecular sieve, the sieve and the mixture forming a single-piece assembly. According to yet another advantageous embodiment, the device comprises a cover formed by a flexible substrate, preferably a metal mesh, the flexible substrate being at least partly covered by a layer formed by the mixture comprising manganese dioxide (MnO2) combined with a compound comprising silver, the compound being an oxide, salt, complex or the like. Thus the package may for example be a drum or a matrix enclosing one or more pieces of radioactive waste. It may also take any other form of container. Advantageously, for each of the three embodiments of the device mentioned above, the mass concentration of manganese dioxide in the mixture is ranging from 80% to 99%, the mass concentration of silver oxide Ag2O in the mixture is between 20% and 1%. This corresponds to a mass fraction of silver (Ag) ranging from 0.93% to 18.6%. This is because 10% Ag2O corresponds to 9.3% silver (Ag). Optionally but nevertheless advantageously, the mixture also comprises a platinum compound the mass concentration of which is preferably ranging from 0.1% to 1%. This corresponds to a mass concentration of platinum ranging from 0.01% to 0.1% for platinum black 10% Pt. An example of an implementation method according to the invention will be detailed. Devices for implementing the method will then be described. In the present patent application, tritiated hydrogen means the gas T2 or HT. Thus tritiated hydrogen differs from tritiated water HTO or T2O, which has a liquid state and a gaseous state. In the latter state, water will be referred to as tritiated water in the gaseous state during the present description. In the context of the present invention, tritiated hydrogen (HT or T2) is oxidized by a specific metal oxide: manganese dioxide (MnO2). The reduction of hydrogen and isotopes thereof by manganese dioxide follows Equation (2):MnO2+Q2→MnO+Q2O (2)where Q represents indifferently all the isotopes of hydrogen. In order to obtain an effective reduction of the quantity of tritiated hydrogen, it is particularly advantageous to have high kinetics for this reaction. To this end, manganese dioxide is associated with a compound comprising silver. In the context of the present invention, a compound comprising silver promotes the conversion reaction. It may for example be an oxide or a salt without this being limitative. Surprisingly, the combination of silver and manganese atoms procures a synergy effect for catalysing the reduction of tritiated hydrogen by manganese dioxide. In addition, the use of manganese dioxide makes the reaction particularly safe and inexpensive, which allows industrial application without the provision of external energy and without maintenance. In the invention, the mixture of silver and manganese oxide is combined with a dehumidifier, typically in the form of a molecular sieve. Advantageously, the molecular sieve is a 4A-type or 5A-type zeolite. Preferably, the package is in contact with a single molecular sieve. The molecular sieve traps the tritiated water. Thus tritiated hydrogen has its mobility reduced when changing in the liquid or gaseous state into tritiated water by virtue of the reaction with the oxide mixture and has its mobility further reduced under the effect of the molecular sieve. The invention therefore makes it possible to transform tritiated hydrogen into tritiated water, tritiated water that is subsequently trapped by the molecular sieve. Moreover, the invention has high efficiency even for small quantities of tritiated hydrogen. However, in the context of degassing of tritiated waste, the rate of degassing of tritiated hydrogen is often very low, typically less than 2 GBq/year/kg of waste or 5.6 10−7 g of tritium/year/kg of waste or 1.8 10−7 mol of tritium/year/kg of waste. The method according to the invention has proved to be particularly effective for treating tritiated waste. Indeed, the method according to the invention has a high reaction kinetics and great reactivity vis-a-vis tritiated hydrogen. In addition, the reactions of the method do not require external maintenance and are not very or not at all reversible, particularly for temperatures below 100° C. Moreover, the reagents such as the products generated by the method are thermally stable. Several ways of preparing the mixture of manganese and silver oxides can be envisaged. A first method consists of mechanically mixing the manganese and silver oxides. More precisely, the manganese dioxide and the silver oxide are first of all produced separately, each of them in the form of a powder. The two powders are then mixed. This method has the advantage of being particularly simple to implement. According to a variant of this first method, water can be added to the manganese dioxide and silver oxide powders. The water is preferentially added in small quantities. This addition of water has the advantage of facilitating the spreading of the silver on the surface of the manganese oxide. Advantageously, 1 ml (10−3 l) of water is added for 2.5 g of mixture of oxides. With this proportion of water, very good homogeneity of the deposition of silver oxide is ensured while limiting the time necessary for evaporation. A second method consists of carrying out impregnation dry. According to this method, a manganese dioxide substrate, preferably in the form of powder or pellets, is wetted with a saline solution including silver ions (Ag+). This saline solution may comprise any silver halide, for example Cl or Br, or silver nitrate. This method is particularly simply to implement. A third method consists of a diffusional impregnation. According to this method, the manganese dioxide substrate is placed in a solution comprising a salt containing silver ions (Ag+) in excess. This method is particularly effective for obtaining a homogeneous distribution of silver. A fourth method is based on the deposition of silver by a precipitation reaction. More precisely, the silver is deposited by precipitation on the manganese using a solution comprising a suspension of manganese oxides and silver nitrate. The precipitation of the silver is obtained by introducing sodium hydroxide (or soda) (NaOH) in the solution. Preferably, the mixing is carried out so that the manganese dioxide and the silver oxide Ag2O have mass concentrations ranging from 80% to 90% and ranging from 20% to 1%, respectively, that is to say a mass proportion of silver of between 0.93% and 18.6%. According to a favoured embodiment, the mass concentration of manganese dioxide is around 90% and the mass concentration of silver oxide is around 10%. FIG. 1 illustrates partial pressure curves of tritiated water (curves 100, 101) and tritiated hydrogen (102, 103) as a function of time according to the method of the invention. To carry out these experiments, the mass concentrations of manganese dioxide and silver oxide Ag2O forming the mixture are 90% and 10% respectively. The mass used of mixture is 0.403 g. The molecular sieve has a mass of 11.241 g. It is formed by 5A-type zeolite. The flow rate of gas containing 133 ppmV of tritiated hydrogen, for example generated by tritiated waste, is approximately 750 Normal milliliters per minute (Nml/min). These curves clearly show the efficiency of the invention in order to reduce the quantity of tritiated hydrogen. The origin of the X axes of the graph corresponds to the moment when the tritiated hydrogen and the mixture of manganese and silver oxides coupled with the molecular sieve are put in contact. The start of the reaction is indicated by the reference 110 and is presented in enlarged form on the graph 111. During the first two minutes approximately, the partial pressure 103 of tritiated hydrogen is relatively stable around a value close to 1.4×10−9 torr, which represents 133 volumetric parts per million (Vppm) of tritiated hydrogen. During this same period, the partial pressure of tritiated water 101 is stable at approximately 4.10−10 torr, which represents an absence of tritiated water. As soon as the tritiated hydrogen is put in contact with the mixture of oxides combined with the molecular sieve, a fast reduction of the partial pressure of tritiated hydrogen is observed, followed by a huge increase of the partial pressure of tritiated water. This represents the production of tritiated water by reaction of the tritiated hydrogen with the mixture of manganese oxide promoted by the silver. Almost instantaneously, the partial pressure of tritiated water 100 declines very quickly. After thirty minutes, it reaches once again a partial pressure representing an absence of water. Simultaneously, the partial pressure of tritiated hydrogen 102 increases in order to reach 8.10−10 torr, corresponding to 75 Vppm after approximately thirty minutes. The change in these curves reflects the action of the molecular sieve acting as a dehumidifier that traps the tritiated water produced by the reaction of tritiated hydrogen with the mixture of oxides. These curves show that the trapping of the tritiated water takes place more quickly than the production thereof. The slow increase of the partial pressure of tritiated hydrogen represents the progressive saturation of the mixture of oxides. When the mixture is saturated, the quantity of tritiated hydrogen increases progressively because of the degassing produced by the waste, whereas the quantity of tritiated water for its part remains zero, the oxidation reaction being interrupted because of the lack of reagent. The calculation of the global yield of the trapping of tritiated hydrogen leads to a reactivity of 23 cm3 of tritiated hydrogen per gram of oxide mixture. The curve in FIG. 1 thus clearly reflects the efficiency of the method according to the invention even when a small quantity of tritiated hydrogen is present. Advantageously, in the context of the limitation of the degassing of packages of waste, this trapping of tritiated hydrogen is carried out under ambient pressure and temperature, which simplifies its implementation thereof. To check the limits of the efficiency of the present invention, the tests were carried out under degassing conditions equivalent to the degassing of 9000 highly degassing packages with in addition a very low mixture mass. This explains why the saturation times are low. Non-limitatively but nevertheless advantageously, this efficiency is further improved by adding platinum (Pt) to the mixture of oxides. Preferably, a platinum compound is added to the mixture of manganese dioxide and silver oxide in small proportions. Typically, the mass concentration of platinum compound is ranging from 0.1% to 1%. This corresponds to a mass concentration of platinum ranging from 0.01% to 0.1% for platinum black 10% Pt. According to a particularly effective example, the platinum black 10% Pt has a mass concentration of 0.56%, the manganese dioxide has a mass concentration of 89.28% and the silver oxide has a mass concentration of 10.16%. FIG. 2 illustrates the results of another example of a method of implementing the invention in which the mixture comprising the manganese dioxide with silver oxide also comprises platinum or a platinum compound. The other conditions for producing these curves are identical to those illustrated in FIG. 1. To carry out these experiments, the mass concentrations of manganese oxide, silver oxide and platinum in the mixture are 89.5%, 10% and 0.5%, respectively. The mass of mixture used is 0.119 g. The molecular sieve has a mass of 9.202 g. As in the previous example, it is formed by a 5A-type zeolite. The flow rate of gas containing 133 Vppm of tritiated hydrogen is also 750 Nml/min. As it can be seen on the curves in FIG. 2, the partial pressure 103 of tritiated hydrogen is relatively stable around a value close to 5×10−10 torr, which represents 12 Vppm of tritiated hydrogen. Next, the saturation of the reagents occurs rapidly because of the small quantity of oxides. The concentration of tritiated hydrogen reaches 85% of its initial concentration, that is to say 115 Vppm after 30 minutes, and reaches its initial value after 90 minutes. The change in the concentration of tritiated water is similar to that illustrated in FIG. 1. The calculation of the global yield of the trapping of tritiated hydrogen leads in the second case to a reactivity of 30 cm3 of tritiated hydrogen per gram of oxide mixture. This represents an increase in the yield of 25% in comparison with the method in which the oxide mixture does not include platinum. Introducing platinum into the mixture therefore has a particularly advantageous effect. As detailed above, the invention is based on the coupling of a molecular sieve with a mixture comprising manganese dioxide and silver oxide. This coupling may appear under various forms. According to a first embodiment illustrated in FIG. 3, the tritiated waste 5 is placed in a drum 10. The drum 10 may be in various forms and in particular in the form of a cylinder having vertical walls 11, a bottom 12 and a lid 13. The drum 10 thus defines a housing being able to contain tritiated waste 5. Preferably, the molecular sieve 1 is placed at the bottom 12 of the drum 10. The waste 5 can then be placed at the bottom of the drum, at least partly covering the sieve 1. The mixture 2 comprising manganese dioxide and silver oxide is then deposited on the surface of the waste 5. The combination of the drum 10, the molecular sieve 1 and the mixture of oxides 2 thus forms a self-contained device making it possible to accept radioactive waste and effectively and durably providing the trapping of tritiated hydrogen and tritiated water. According to another embodiment, the mixture comprising manganese dioxide and silver oxide is deposited on the sieve. Preferably this deposition forms a single-piece assembly. It is then easy to handle. In this case, provision can be made for fixing the mixture 2 on the sieve 1 mechanically by stirring the species put in contact. Alternatively or accumulatively, this fixing can be provided or reinforced by means of an additive. It is possible for example to use water as an additive. Therefore this water is evaporated by drying the mixture of oxides between 150° and 200° C. for 12 h and 48 h. The mixture 2 thus forms a layer 3 partly covering at least the sieve 1. The sieve 1 is covered so as not to impair the dehumidification properties of the sieve 1. The assembly formed by the molecular sieve associated with the mixture so as to form a single-piece assembly is in itself particularly advantageous since it can be put in contact with tritiated waste under varied forms and packages. According to a particular embodiment illustrated in FIG. 4, the assembly formed by the sieve 1 and mixture 2 is placed in a drum 10, for example at the bottom 12 of the drum 10. It then suffices to place the waste inside the drum 10 so that the trapping of tritiated hydrogen takes place along with the degassing. According to another embodiment, the molecular sieve is placed at any point in the drum. According to another embodiment, the radioactive waste is coated in a matrix. The package is thus formed from the matrix and waste. This matrix is for example made from cement, bitumen or glass. In this case, it is advantageous to fix the mixture comprising the manganese and silver oxides on a flexible structure such a metal mesh so as to form a cover for enclosing the waste. The cover is preferably flexible. Preferably, the molecular sieve is integrated to the cover. According to the above description, it is clear that the invention affords an effective response to the problem of the degassing of tritiated hydrogen and tritiated water. It is expected that, during the whole of its operating life as well as its dismantling period, a reactor of the ITER type will produce approximately 35,000 tons of radioactive waste. Among this waste, pure tritiated waste and very weakly radioactive waste will have to be placed directly in drums. In order to comply with the storage constraints, the degassing of tritiated hydrogen should be around 0.1 milligrams per year and per drum. Thus, with this theoretical degassing rate and even though the efficiency of the trapping of the tritiated hydrogen would be low (an efficiency of 0.3%), a single gram of mixture comprising manganese dioxide combined with silver oxide would be necessary. With a degassing rate of 0.1 milligrams per year and per drum, a molecular sieve of 120 g of zeolite per year and per m3 of drum should also be sufficient to provide trapping of tritiated water present in the vapour state and in the form of liquid tritiated water. This relatively small quantity of materials representing the molecular sieve should also prove to be sufficient even with extreme conditions of heat (at least up to 40° C.) and humidity (at least 40%). While the invention has been shown and described with reference to a certain embodiments thereof, it will be understood by those skilled in the art that various changes in form and detail may be made therein without departing from the spirit and scope of the invention as defined by the appended claims and their equivalents.
	summary
056235267	claims	1. A method of repairing a nuclear reactor which includes a plurality of fuel rods and a shroud which is disposed about said fuel rods comprising the steps of: placing a strap against the external surface of a nuclear reactor shroud so as to span at least one weld formed in said shroud; forming holes at each end of said strap and in said shroud so that a hole in said strap is precisely aligned with a hole formed in said shroud, and in a manner wherein at least one of the holes which are aligned is formed in situ; and securing said strap to said shroud by inserting fastening means into said holes. placing an elongate strap against a predetermined portion of an external surface of the shroud wherein the strap spans at least one weld which secures segments of said shroud together; machining, in situ, at least one of a matched pair of holes which respectively extend through the strap and the shroud; and disposing expandable bolt units in each of said matched pair of holes and using said bolt units to fasten said straps to said shroud to render said straps rigid with the shroud. a strap which is formed of the same material as said shroud, which extends essentially parallel to an axis of said cylindrically-shaped shroud, and which is securely fastened at its both ends to an external surface of the shroud, said strap being arranged to span at least one segment and a plurality of welds which interconnect the annular segments of said shroud, said strap withstanding bending, shearing and tension forces which are applied to said shroud; and a plurality of predetermined bolt units which are adapted to be inserted from an exterior of said shroud into pairs of aligned holes which are located at each end of said strap, and at least one of which is formed in situ by a cutting device, each of said bolts units having a portion which can be expanded in a manner which secures the bolt unit in the hole in which it is inserted, wherein each end of said strap is provided with a structure for locating an EDM cutting device in a manner wherein the hole which is formed in said shroud is precisely aligned with a hole in said strap. means for placing a strap against the external surface of a nuclear reactor shroud so as to span at least one weld formed in said shroud; means at each end of said strap for forming precisely aligned holes in said strap and said shroud; and means for securing each end of said strap to said shroud by inserting fastening means into the precisely aligned holes. means for placing an elongate strap against a predetermined portion of an external surface of the shroud wherein the strap spans at least one welds which secure segments of said shroud together; means for machining, in situ, at least one of a matched pair of holes which each extend through both the strap and the shroud; and means for disposing expandable bolt units in each of said plurality of holes and using said bolt units to fasten said straps to said shroud and render said straps rigid with the shroud. 2. A method as set forth in claim 1, wherein said fastening means comprise bolt units which have a portion which can be expanded in a manner which secures the bolt unit in the holes in which it is disposed. 3. A method as set forth in claim 1, wherein said shroud is comprised of a number of segments which are welded together and wherein said at least one weld formed in said shroud is a weld which interconnects two of said segments. 4. A method as set forth in claim 1, further comprising the step of forming each end of the strap with a structure for supporting hole making equipment in an operative position thereon. 5. A method of repairing a nuclear reactor which includes a plurality of fuel rods and a shroud which is disposed about said fuel rods, comprising the steps of: 6. In a nuclear reactor having an essentially cylindrically-shaped shroud formed of a plurality of annular segments which are welded to one another: 7. An apparatus for repairing a nuclear reactor which includes a plurality of fuel rods and a shroud which is disposed about said fuel rods comprising: 8. An apparatus as set forth in claim 7, wherein said fastening means comprises a bolt which has a portion which can be expanded in a manner which secures the bolt unit in the aligned holes in which it is disposed. 9. An apparatus for repairing a nuclear reactor which includes a plurality of fuel rods and a shroud which is disposed about said fuel rods, comprising:
	claims	1. A modular assembly comprising:a base member;a stage member movable relative to the base member;a filter-foil device coupled to and thereby moveable with the stage member, said filter-foil device comprising a body, one or more photon flattening filters and one or more electron scattering foils supported by the body, the body being movable relative to the stage member in positioning the photon flattening filters or electron scattering foils; anda target device supported by the base member, said target device comprising a substrate and one or more targets supported by the substrate, the substrate being movable in positioning the one or more targets. 2. The modular assembly of claim 1 which comprises:a first driving device supported by the base member, the first driving device being operable to move the stage member and the filter-foil device coupled to the stage member relative to the base member;a second driving device supported by the stage member, the second driving device being operable to move the body of the filter-foil device; anda third driving device supported by the base member, the third driving device being operable to move the substrate supporting the one or more targets of the target device. 3. The modular assembly of claim 2 which further comprises:an ion chamber device; anda fourth driving device supported by the base member, the fourth driving device being operable to move the ion chamber device relative to the base member. 4. The modular assembly of claim 3 which further comprises a field light device, the field light device comprising a mirror member supported by and thereby being movable with the body of the filter-foil device, and one or more light sources supported by and thereby movable with the ion chamber device. 5. The modular assembly of claim 3 wherein at least one of the first, second, third, and fourth driving devices comprises a servo motor and one or more feedback devices. 6. The modular assembly of claim 3 wherein each of the first, second, third, and fourth driving devices comprises a servo motor and one or more feedback devices. 7. The modular assembly of claim 1 wherein the one or more photon flattening filters are arranged in an arc or a circular configuration having a first radius, and the one or more electron scattering foils are arranged in an arc or a circular configuration having a second radius that is different from the first radius. 8. The modular assembly of claim 1 wherein the stage member is movable relative to the base member in a linear direction, and the body of the filter-foil device is movable at least in a rotary direction. 9. A system comprising a beam filter positioning device and a controller operable to control the beam filter positioning device using a computer software embodied in a non-transitory computer readable medium, wherein the beam filter positioning device comprises:a base member;a stage member movable relative to the base member;a filter-foil device coupled to and thereby moveable with the stage member, said filter-foil device comprising a body, one or more photon flattening filters and one or more electron scattering foils supported by the body;a first driving device supported by the base member and being operable to move the stage member and thereby the filter-foil device coupled to the stage member relative to the base member; anda second driving device supported by the stage member and being operable to move the body of the filter-foil device and thereby the photon flattening filters and electron scattering foils supported by the body relative to the stage member; andwherein the first and the second driving devices are controllable by the controller using the computer software. 10. The system of claim 9 wherein each of the first and second driving devices comprises a servo motor and one or more feedback devices which are coupled to the controller, and the controller is operable to control the servo motor based on at least signals from the one or more feedback devices. 11. The system of claim 10 wherein the beam-filter positioning device further comprises:a target device supported by the base member, the target device comprising a substrate and one or more targets supported by the substrate; anda third driving device supported by the base member and operable to move the substrate and thereby the one or more targets supported by the substrate;wherein the third driving device comprises a servo motor and one or more feedback devices which are coupled to the controller, and the controller is operable to control the servo motor of the third driving device based on at least signals from the one or more feedback devices of the third driving device. 12. The system of claim 11 wherein the beam-filter positioning device further comprises:an ion chamber device; anda fourth driving device supported by the base member, the fourth driving device being operable to move the ion chamber device relative to the base member;wherein the fourth driving device comprises a servo motor and one or more feedback devices which are coupled to the controller, and the controller is operable to control the motor of the fourth driving device based on at least signals from the one or more feedback devices of the fourth driving device. 13. The system of claim 12 which further comprises a field light device, the field light device comprising a mirror member supported by and thereby being movable with the body of the filter-foil device, and one or more light sources supported by and thereby movable with the ion chamber device. 14. The system of claim 12 wherein the controller is operable to control:the first driving device to move the stage member in a linear direction;the second driving device to move the body of the filter-foil device in a rotary direction;the third driving device to move the substrate in a linear direction; andthe fourth driving device to move the ion chamber device in a linear direction. 15. The system of claim 9 wherein the one or more photon flattening filters are arranged in an arc or a circular configuration having a first radius, and the one or more electron scattering foils are arranged in an arc or a circular configuration having a second radius that is different from the first radius. 16. The system of claim 9 wherein the controller is operable to control:the first driving device to move the stage member in a linear direction;the second driving device to move the body of the filter-foil device in a rotary direction.
050892100	summary	The present invention relates to nuclear reactors and more specifically to a fuel bundle arrangement for a boiling water nuclear reactor in which so-called mixed oxide fuels including plutonium and uranium are utilized in a nuclear fuel bundle together with a burnable absorber such as gadolinium to optimize the reaction of a nuclear fuel bundle. BACKGROUND OF THE INVENTION 1. Field of the Invention In a boiling water nuclear reactor, a fissile fuel atom, such as U235, PU239, or PU241 absorbs a neutron in its nucleus and undergoes nuclear disintegration. This produces on the average two fissile fragments of lower atomic weight with great kinetic energy and several neutrons, these neutrons also being at high energy. In the boiling water nuclear reactor, the nuclear fuel is in the form of fuel rods, each of which comprises a plurality of scintered pellets contained within an elongate sealed cladding tube or "fuel rod." Groups of such fuel rods are supported between upper and lower tie plates to form separately replaceable fuel assembles or bundles. A sufficient number of such fuel assemblies are arranged in a matrix, approximating a right circular cylinder, to form the nuclear reactor core which is capable of self-sustained fission reaction. The kinetic energy of the fission products is dissipated as heat in the fuel rods. Energy is also deposited in the fuel structure and moderator by the neutrons, gamma rays and other radiation resulting from the fission process. The core is submerged in coolant which removes the heat. Typically such heat removal occurs by the coolant water boiling into steam. From the steam energy is extracted to perform useful work. In a boiling water nuclear reactor the coolant also acts as a neutron moderator. This moderator takes the emitted high energy neutrons and slows down the neutrons to render them thermal in character and hence more likely to be absorbed in the fuel continuing the fission reaction. The commonly used fuel for water cooled and moderated nuclear power reactors comprises uranium dioxide of which from about 0.7 to 5.0% is fissile U235 mixed with fertile U238. During operation of the reactor, some of the fertile U238 is converted to fissile plutonium PU-239 and PU-241. It turns out that the U238 is also fissionable, but only for high energy neutrons. The ratio of fissile materials produced (for example plutonium 239 and plutonium 241) to fissile material destroyed (for example U235, plutonium 239 and plutonium 241) is defined as the "conversion ratio". Fuel bundles are typically replaced at certain "outages". Typically these outages delineate "cycles". During such outages the reactor is opened and remote lifting equipment removes fuel bundles with spent fuel therein and replaces the fuel bundles with those having fresh fissile materials contained within the fuel rods. At each outage, only a portion of the total fuel bundles are removed. This portion is in the order of 25%. STATEMENT OF THE PROBLEMS It is known in the prior art how to reprocess the fuel from used fuel bundles. Typically, plutonium is recovered (for example PU239, PU241). This recovered plutonium is mixed by taking naturally occurring uranium oxides or diffusion plant tails uranium and undergoing a blending process. Typically, the plutoniums and uraniums are converted to oxides. Thereafter, the plutonium oxides and uranium oxides are blended to produce a desired percent by weight of the respective plutonium and uranium compounds. Once the oxides are blended, they are scintered to generate preferably a face centered cubic lattice structure incorporated in fuel pellets. The scintered pellets are placed in sealed zirconium fuel rods. Regarding the matter of blending, these mixed oxide or MOx fuel rods are required to have varying concentrations of Pu. This being the case, it is often necessary to have five or six differing concentrations of plutonium in the varying rods of an individual fuel bundle. It can be appreciated that this sort of recovery, blending and scintering of the plutonium from spent or used fuel bundles to new fuel bundles is difficult. The radioactive gases and elements of the spent fuel rods must be handled, typically remotely. The plutonium itself is toxic, is an alpha emitter and has long half-lives which complicate the removal process. Simply stated, such plutonium recovery, blending and scintering operations must occur in enclosed environments under special processing conditions. Accordingly, the processing and placement of mixed oxide fuels within nuclear fuel rods differs radically from the use of the more conventional uranium oxide compounds. This being the case, the use of so-called mixed oxide fuels (MOx) occurs in isolated processing plants separate and apart from conventional fuel loadings involving uranium oxides and burnable absorbers such as gadolinium. If the reactor is to operate at a steady state power level, the fission inducing neutron population must remain constant. That is, each fission reaction must produce a net of 1 neutron which produces a subsequent fission reaction so that the operation is self-sustaining. The operation is characterized by an effective multiplication factor K.sub.eff which must be at unity for a steady state operation. It is noted that the effective multiplication factor K.sub.eff is the neutron reproduction factor of the nuclear reactor considered as a whole. This is to be distinguished from the local or infinite multiplication factor K.sub.inf which defines neutron reproduction of an infinitely large system having throughout the same composition and characteristics as the local region of a reactor core in question. During operation, the fissile fuel is depleted and indeed some of the fission products are themselves neutron absorbers or "poisons". To offset this, the reactor must normally be provided with an initial excess of nuclear fuel which results in initial excess reactivity. This initial excess reactivity requires a control system to maintain the effective multiplication factor at unity. Such maintenance must occur at the beginning of the fuel cycle in the presence of the excess reactivity. The control system also functions to reduce the effective multiplication factor to below unity so that the reactor can be shut down. The control system typically utilizes neutron absorbing material which serves to control the neutron population by nonfission absorption or the capture of neutrons. As the described invention herein relates to a sophisticated manner of loading a fuel bundle, it is necessary to understand both the construction of a typical fuel bundle together with its geometry in relating to a so-called control rod as well as adjacent fuel bundles. Each fuel bundle contains longitudinally extending sealed rods. These rods have the fissile material sealed within them. The material is maintained sealed during the entire life of the fuel bundle. Opening the sealed fuel rods must occur for the processing referred to here. A group of such fuel rods are typically supported between a lower tie plate at the bottom and an upper tie plate at the top. Arrays of 6.times.6, 7.times.7, 8.times.8 and 9.times.9 fuel rods have been utilized. Typically, these arrays define in the indicated number certain so-called "lattice positions" for the fuel rods. It will be understood that portions of these lattice positions can be occupied by other fuel bundle elements. For example, it is common to insert rods for holding water moderator in the center of such fuel bundles to impart to the entirety of the bundle the desired reaction profile for the generation of energy. Typically, the total number of fuel rods has increased with modern fuel bundle design. Presently, lattice positions 8.times.8 and 9.times.9 are utilized. Furthermore, in the 9.times.9 designs, it is now common to have foreshortened fuel rods in some of the lattice positions. Foreshortening of the fuel rods imparts numerous advantages. An example of such advantages is set forth in the patent application entitled Two Phase Pressure Drop Boiling Water Reactor Assembly Design Ser. No. 176,975 filed Apr. 4, 1988. Each fuel bundle includes a surrounding channel for confining water flow between and through the tie plates on the axial length of the individual rods. Water moderator flows in the confining channel from the bottom and out of the confining channel at the top. During its passage in an active reactor, steam is produced in mixture with the passing water. The water moderator is also exterior of the channel. This water typically does not include a high percentage of steam and is contained within what is known as a core bypass zone or region. The core bypass zone or region has water which produces moderation of the fast neutron flux. Fast neutrons rapidly become thermal neutrons capable of initiating continuing nuclear reaction within the core bypass region. The fuel bundles themselves are typically arrayed for controlling their nuclear reaction in groups of four. Typically, four fuel bundles of square cross section are vertically disposed in side by side relation. Each fuel bundle is spaced apart from the remaining fuel bundle so as to define the interstitial core bypass region. Considering four fuel bundles in side by side relation, the bundles will in the interstices between them define a cruciform shape interstitial area. It is into this cruciform shape interstitial area that a complementary cruciform shaped control rod effects penetration for the parasitic absorption of neutrons and ultimate control of the nuclear reactor. Modern control rod design includes many groups of four such fuel bundles. It is common for reactor cores to hold up to 800 such discrete fuel bundles. Each group of four such fuel bundles has a control rod penetrating interstitially of the fuel bundle interfaces to affect the absorption of neutrons and the control of the nuclear reactor. During the lifetime of a fuel bundle, the usual circumstance is that the bundle is not appreciably exposed to the control rods. Consequently, it is the usual case that the thermal neutron flux is relatively high at the fuel bundle corners in comparison to other portions of the fuel bundle. As the fuel bundles are initially supplied with excess reactivity within their fissile materials it is sometimes necessary to incorporate burnable absorbers. Such burnable absorbers act during the beginning of a fuel cycle to absorb neutrons and prevent the excess reactivity which would otherwise be present from preventing control of the nuclear reaction. A burnable absorber is a neutron absorber which is converted by neutron absorption into a material of lesser neutron absorbing capability. A well known burnable absorber is gadolinium. The odd isotopes (GD-155 and GD-157) have very high capture cross sections for thermal neutrons. The burnable absorbers available for use also have an undesirable effect. Specifically, and during the end of the fuel bundle cycle, the residual burnable absorbers decrease the efficiency of the fuel bundle. Its operation would be far better if the burnable absorbers were not present, or at least maintained at an absolute minimum. For example, if gadolinium is used as a burnable absorber, the high cross section isotopes (GD-155 and GD-157) deplete rapidly. Unfortunately, these elements are converted into elements which contain reduced neutron capture cross sections but nevertheless detract measurably from the efficiency of the fuel bundle. For example, in gadolinium, the produced isotopes (GD-154, GD-156 and GD-158) still continue to absorb neutrons and detract from the overall efficiency of the fuel bundle. As is well known, burnable absorbers, such as gadolinium operate in a self-shielding mode when present at sufficient concentration. That is, upon exposure to neutron flux, the neutron absorptions occurs essentially at the outer surface of the absorbers so that the volume of absorber shrinks radially at a rate that depends upon the concentration of the absorber. It is additionally known that plutonium, especially fissile PU239 and PU241 have high neutron absorption cross sections relative to uranium. If burnable absorbers such as gadolinium are utilized in combination with fissile plutonium, the gadolinium itself can be shielded from neutrons by the plutonium. Hence to use the control feature of the burnable absorber, much larger concentrations of gadolinium must be utilized where fissile plutonium is present. There is, related to the present fuel design, a further complication. During operation, the percentage of steam voids within the fuel bundles increases to and towards the top of the reactor. These steam voids lead to decreasing moderation in the top regions of the reactor because water is present in lesser quantities. Thus, there results a power distribution that is skewed towards the lower regions of the fuel bundles forming the reactor core. It is a known practice to compensate for this by distributing burnable absorber in an axially inhomogeneous manner. A number of fuel rods are provided with burnable absorber having a distribution skewed towards the axial region of hot operating maximum reactivity. A typical configuration is shown in U.S. Pat. No. 3,799,839. However, the situation is very different in the cold shutdown state. More particularly, in the cold state, the top of an irradiated boiling water reactor core is more reactive than the bottom. This greater reactivity occurs because during normal operation there is greater plutonium production and less U235 destruction in the reactor top. Specifically, a greater population of fast neutrons is present at the top of the reactor. These fast neutrons create a greater conversion ratio and smaller burnup in the fuel rods. In the cold shutdown condition, the steam voids in the upper part of the core are eliminated because little, if any, steam is present in the moderator. This makes the top of the core more reactive than the bottom in the cold shutdown condition. Typical licensing standards require a 0.38% reactivity shutdown margin (K.sub.eff less than 0.9962) with any one control rod stuck out of the core. To provide margin for prediction uncertainties, a design basis of 1% predicted shutdown margin (K.sub.eff less than 0.99), to be provided by the control rods and the burnable absorber is typically required and used. Some reactors have fuel assemblies requiring special designs directed to their so-called "cold reactivity." In such prior art fuel assemblies, the burnable absorber is asymmetrically distributed to allow cold shutdown margins to be met with minimum penalty to operating efficiency. The assembly includes a component of fissile material distributed over the axial extent of the fuel bundle. Mixed within fissile materials a component of neutron absorbing material is added. This neutron absorbing material has an axial distribution characterized by an enhancement in a relatively short axial zone known as the "cold shutdown control zone". This cold shutdown control zone corresponds to at least a portion of the axial region of cold shutdown maximum reactivity. The axial distribution of the component of neutron absorbing material is typically characterized by an additional enhancement in an axial zone known as the "hot operating" control zone. The component neutron absorbing material is conventionally incorporated into at least some of the fuel rods. This enhancement in the cold shutdown zone may be provided at least in part by one or more fuel rods having absorber only in the cold shutdown control zone. This enhancement of the neutron absorbing material in the cold shutdown control zone may be additionally supplemented by reduced fuel enrichment in the cold shutdown control zone. It should be further understood that it is desired to keep the distribution of gadolinium within a fuel bundle to an absolute minimum. Gadolinium, in addition to absorbing neutrons, reduces the thermal conductivity of the fuel rods and increases fission gas release. Consequently the gadolinium containing rods are frequently the most limiting rods in the fuel assembly. Thus, and because of these limiting rods, the entire fuel bundle must be downrated in power with a corresponding adverse effect on local power distributions. The amount of power downrating that is required depends upon the required gadolinium concentration. This required gadolinium concentration sometimes becomes a serious problem in extended burnup fuel designs and/or high energy cycle designs where increased gadolinium concentrations are required in order to provide adequate cold shutdown margins. Unfortunately, the relatively dense 9.times.9 arrays utilized with modern reactors are examples of fuel bundles in which the excess gadolinium can produce problems. The reader must appreciate at this juncture that the above recited background includes a summary of only pertinent operating and shutdown constraints. These pertinent constraints have been summarized so that the following optimized fuel bundle design can be understood. SUMMARY OF THE INVENTION A fuel bundle design incorporating oxides of recovered plutonium mixed with uranium (MOx) which can maximize the content of plutonium and minimize the number of different MOx pellet concentration types. In a boiling water nuclear reactor, a fuel bundle is loaded with MOx containing rods at all locations save and except rods at the corners or adjacent to the corners of the bundle. Fuel rods adjacent the corners are provided which preferably do not contain MOx and are instead uranium gadolinium rods. The disclosed uranium gadolinium rods can have their gadolinium asymmetrically loaded so as to impart axially of the fuel bundle the desired cold reactivity shutdown zones. As a consequence, a fuel bundle design is disclosed which can maximize the use of recovered fissile plutonium from previous fuel cycles, minimize the number of different MOx concentration type and enable a reduction of the amount of uranium enrichment required. At the same time, the desired axial shaping for the so-called cold reactivity shutdown zones can be accomplished independent of the MOx rods and more importantly without the mixture of MOx fuels and gadolinium in the same fuel rods. Further, by the placement of the gadolinium uranium rods at the periphery of the bundle near the permanent water gaps with their high thermal neutron flux, the maximum worth of the gadolinium is achieved. Shielding of the gadolinium by the high neutron cross section plutonium is minimized. With the disclosed design, uses of the burnable absorber gadolinium is reduced to a minimum resulting in an improved overall fuel bundle energy output. Other Objects, Features and Advantages An object of this invention is to disclose a fuel bundle design in which high levels of plutonium per bundle are utilized. An advantage of the disclosed design is that it limits the number of discrete plutonium concentrations that are required in each MOx fuel bundle. In the preferred embodiment herein disclosed only three variant concentrations of plutonium are required. This simplifies the blending, mixing and individual fuel rod assembly by limiting the number of discrete plutonium mixtures that are used with the disclosed design. A further object of this invention is to realize maximum worth of the gadolinium that is utilized with the disclosed fuel bundle. In accordance with this aspect of this invention the gadolinium is placed in corner locations. In these corner locations, the gadolinium sees relatively high neutron flux. At the same time, the gadolinium is not shielded by the relatively high cross section of plutonium used as the fissile material in the bundle. An advantage of this aspect of the invention is that the gadolinium is disposed where it easily accommodates excess reactivity during the first part of the fuel cycle. A further advantage of this feature is that when the gadolinium is expended, typically after the first quarter of the end reactor life of the fuel bundle, the remaining gadolinium presents a minimal inefficiency to the fuel bundle. Yet another advantage of this invention is that the gadolinium placed within fuel rods only at the corner of the fuel bundle can be utilized for the purpose of imparting the cold reactivity shut down zone to the fuel bundle. No varying of the plutonium concentration of the so-called MOx rods is required. Yet another object of this invention is to disclose a MOx fuel bundle which has a higher reactivity profile during its full life cycle within a reactor. This higher reactivity is imparted to the whole core with the advantage that the higher reactivity helps maintain a fissile reaction. An advantage of this aspect of the invention is that with the disclosed fuel bundles dispersed throughout the core, the requirement for enrichment in neighboring fuel bundles is reduced. A further advantage of this disclosed design is that it can be utilized with the dense array of modern fuel bundle designs. For example, two embodiments here utilized are 8.times.8 and 9.times.9 fuel rod arrays. Yet another advantage of this invention is that the design can accommodate partial length rods. Typically, the partial length rods, being placed inside the bundle, are the MOx rods in the disclosed design.
039980570	summary	BACKGROUND OF THE INVENTION The invention covers a nuclear power plant with a closed system of circulating cooling gas. The equipment consists of a high-temperature reactor, a gas turbine assembly and heat exchangers. The latter are made up of recuperators, pre- and intermediate coolers as well as pipes carrying the gas between the several components. The entire assembly is encased in a housing made of pre-stressed concrete (single-unit construction). Plants of the type described offer obvious advantages over nuclear power installations of a different, existing type in which the energy is transferred to a secondary circuit, because they combine the advantages of gas turbines with the high efficiency and simplicity in operation which is typical of single-circuit construction. By integrating the reactor, the turbines, the required coolers, and all other circuit components in a single pressure tight vessel, separate connecting elements connecting the components that contain "live" fuel are rendered unnecessary, a fact which offers distinct advantages in the construction and operation of high-temperature reactors. The integrated design is therefore preferred for a large number of specialized nuclear power plants. For example, German Auslegeschrift No. 1,156,903 describes a compact power plant of the above-mentioned type intended for vehicles. In this design the turbine and the compressor are located on opposite faces of the reactor core, a hollow shaft passing through the core, with intermediate coolers being located in the annular space between the reactor and the wall of the pressure chamber. In this compact design it is assumed that the turbine used will not require any maintenance, and therefore no provision has been made for the removal of the turbine or any other components of the circuit. A similar construction is shown in a nuclear reactor disclosed in German Offenlegungschrift No. 2,005,208 where a pressure jacket, open at its frontal surfaces, is located inside the pressure chamber and is spaced relative to the inside wall of the pressure vessel so as to accommodate the heat exchangers. Also, the German Offenlegungsschrift No. 2,028,736 a nuclear power plant is described having a system of closed circulation of gas. This power plant is built on the principle of dual chambers. The gas turbine together with the components belonging to the gas circulation system are contained in a block of prestressed concrete separate from the concrete pressure vessel, with the object of simplifying the fueling- and controlling processes. A similar construction is used in the high-pressure concrete vessel described in German Auslegeschrift No. 1,614,610, consisting of two separate pressure-tight chambers, one containing the reactor, the other the secondary equipment. The working medium is carried by pipes protruding through the pressure-tight walls, first from the reactor to the turbine and from the compressor then back into an annular-shaped space below the reactor core. This so-called "igloo-method" however offers technical difficulties in construction, and the nuclear power plant, because of the principles applied in the arrangement of components, is highly uneconomical. German Offenlegungsschrift No. 2,062,934 likewise discloses a gas-cooled nuclear reactor of integrated construction in which the gas turbine is located inside the hollow wall of the pressure chamber enclosing the reactor core. Through a by-pass mechanism a part of the cooling gas carried past the core of the reactor can be diverted and directly combined with the hot exhaust gas emitted from the core. Another nuclear power plant of the type originally defined is described in German Offenlegungsschrift No. 1,764,249. Here the nuclear reactor together and all circulatory components are located in closely spaced, parallel, vertical shafts inside the concrete pressure vessel, with all components being fully accessible from the outside. Passages for the cooling media are provided within the wall of the pressure vessel as well as between the several vertical shafts. In this design the cooling media has to flow over an extensive area, resulting in the need for a relatively bulky pressure vessel in this type plant. SUMMARY OF THE INVENTION The present invention proceeds from the foregoing state of the art and has as an object to correct the shortcomings inherent in the known nuclear power plants through a particular arrangement of all components. It is a particular object of the invention to provide a compact nuclear power plant, while at the same time rendering all components easily accessible. In accomplishing these and other objects, there has been provided in accordance with the present invention a nuclear power plant comprising an inner, generally cylindrical vessel; a high-temperature reactor contained within the inner vessel; a gas turbine assembly located in a horizontally oriented chamber positioned in the inner vessel beneath the reactor; a plurality of heat exchanger means positioned in a plurality of vertically oriented pods spaced radially, preferably in a circle about the reactor in the inner vessel; and conduit means interconnecting the reactor, the turbine assembly and the heat exchanger means for carrying a gas between these components. The conduit means are arranged in essentially horizontal and vertical straight lines, and the conduit means connecting the reactor and the turbine for carrying high pressure gas are provided with horizontal connections to the reactor and the turbine. The conduit means for carrying low pressure gas comprises a horizontal conduit system positioned beneath the turbine assembly and is comprised of a plurality of coaxial connecting tubes, collectors and distributors and a plurality of normal conduits. The remainder of the heat exchange means comprise intermediate coolers, preferably arranged in vertically stacked pairs in each pod, and coaxial conduits are provided for transporting gas from the low-pressure compressor to the intermediate coolers, through the outer passage of the conduit, and back from the coolers to the high-pressure compressor through the inner passage of the conduit. A plurality of secondary heat absorption devices are preferably also located in additional vertical pods arranged around the reactor. The structural features of the present invention may thus be summarized as follows: The gas-turbine system is installed in a horizontal pod located underneath the reactor which is located centrally inside a vessel; several vertical pods, placed symmetrically around the reactor, contain the recuperators and the pre-coolers; the pre-coolers serving the recuperators are placed above or below the recuperators; the tubes carrying the gas between components run in a straight path, either vertically or horizontally, with vertical tubes being in pods; the high-pressure circuit runs through several vertical pods equipped with horizontally placed connections; a horizontally positioned system of tubes carrying low-pressure gases is placed underneath the turbine compartment. The principles realized in the nuclear power plant according to the invention may be briefly summarized as follows: a largely symmetrical structure of the vessel made of pre-stressed concrete; development of pods as vertical gas lines; gas lines connecting the various components in the primary circuit are direct and in straight lines; gas lines are arranged coaxially and maintain in normal operation only minor differences in pressure between the gas streams flowing coaxially; streams of hot gases flow coaxially inside an isolated system of tubes that are freely distributed inside the pre-stressed concrete vessel and shielded by high-pressure jackets which are surrounded by a circuit of cold gas; easy accessibility from the outside of secondary equipment built into the structure, such as gas lines, heat exchangers, values, heat insulators, etc. for the purpose of inspection, maintenance, repair, and removal after shutting off certain sections of the plant. All pods, gas lines and components of the primary circuit are geometrically accessible upon the removal of the lid of the concrete container, allowing for inspection, maintenance and repair by remote control. The convenient accessibility is the result of the coaxial arrangement of gas lines in relatively large concrete channels, the direct gas leads being arranged in straight lines, and the use of pods serving as vertical gas ducts. By locating two separate gas circuits in one single concrete unit, compactness of the primary circuit is achieved, and the dimensions of the concrete container can be held down to a relatively small size. The arrangement of the various components is such that it can be retained at any desired level of varying output without difficulty, a fact which is of great importance in the development of new types of nuclear reactors. The turbine assembly is built in a horizontal chamber placed at a distance from the nuclear reactor which offers adequate shielding to the turbine assembly against neutron radiation. For the installation and removal of the turbine assembly a sliding device has been provided. The turbine assembly is constructed with a single shaft, offering distinct advantages over a multiple-shaft design: its operation and normal functioning are easy to supervise and of proven reliability; it requires only a single seal where the shaft penetrates through the concrete jacket; and its cost is lower. The turbine is rigidly coupled to the generator. The hot gas coming from the reactor is first taken up by four radially located connecting pipes and then carried over vertical ducts containing hot gas, designed as pods. Then, it is carried by four horizontal connecting tubes to the symmetrically designed intake tubes of the turbine. The four radial connecting pipes together with their graphite packings extend all the way to the vertical gas pipes formed as pods. The exhaust gas from the turbine (approx. 500.degree. C.) first flows downward in a vertical pipe where it enters into the horizontal system of tubes carrying low-pressure gas. By distributors and coaxial cross-connections or feeder lines the gas is distributed to the pods which contain the recuperators and pre-coolers. Upon entering the recuperators it flows through them along the jacket side. The gas subsequently passes through the pre-coolers and is finally returned to the horizontal pipe system. In the return flow, the gas is directed through the outer passages if the coaxial ducts, while on its way to the recuperators it is directed through the interior passages of the coaxial feeder line system. The coaxial network of cross-connections or feeder lines is designed in such a manner that the exhaust gases leaving the turbine at about 500.degree. C. are encircled on all sides by cold gas, whereby thermal stresses in the gas lines are minimized. The cold gas (30.degree. C.) from the horizontal pipe system enters a vertical cylindrical duct through several simple horizontal ducts and is then brought into the low-pressure compartment of the compressor. Here it is compressed to 36 bar. In an advantageous embodiment of the invention six recuperators are provided, which are connected in pairs via a coaxial feeder line to a collector and a distributor. Accompanying pre-coolers are located respectively vertically underneath the recuperators. The compressed gas is carried back to the recuperators by six pipes. A major portion of four of these six pipes runs coaxially to the vertical gas pipes leading from the reactor to the turbine. Thereby the relatively cold high-pressure gas (125.degree. C.) encircles the four hot-gas pipes between the reactor and the turbine. In order to increase the efficiency of the nuclear power plant, an intermediate cooling system is provided in the primary circuit, also located in the vertical shafts or pods. These pods are arranged in the same circle about the reactor as the pods holding the recuperators and pre-coolers and are symmetrically placed with respect thereto. The intermediate coolers are connected in two groups, each containing a pair of intermediate coolers, installed on top of each other in a single pod. Two coaxial pipes lead from the low-pressure compartment of the compressor to the two pods holding the pair of intermediate coolers. The gas flows in the outer pipes toward the pods where it is divided into two partial streams, one flowing upward, the other downward. After passing through the intermediate coolers the gas is carried back to the high-pressure compartment of the compressor through the interior tube of the coaxial system. The hollow pods in the concrete housing, designed to house the components, such as the reactor, the horizontal turbine compartment, vertical pods for heat exchangers, gas ducts and regulators, are preferably lined with gas-tight steel liners. Excess pressures are taken up by the concrete jacket and to reduce the build-up of excessively high temperatures in the concrete, the liners are watercooled and further protected by insulation. As mentioned before, the intermediate coolers serve to increase the efficiency of the plant. It is, however, conceivable to design nuclear power plants in which intermediate coolers are omitted, whereby a reduction in efficiency may be accepted in exchange for various other advantages. The most important of these advantages may be summarized briefly as follows: smaller dimensions of the pre-stressed concrete vessel; elimination of expensive components (besides the intermediate cooling units also steel liners, gas ducts, and other devices connected with the construction of the installation); a reduction in the size of the cooling system; and a reduction of pressure-losses connected with circulating systems. In a power plant of this type the stream of gas leaving the compressor is conveyed directly into the recuperators. It is of advantage to additionally provide a final stage for the elimination of heat inside the stressed-concrete casing, conventionally consisting of a blower, equipped either with or without a recuperator, and a cooler. Such a cooling system operates independently of the main circulation system for the described single-shafted gas turbine assembly; it provides for the disposal of the reactor's heat in the event of a turbine failure, in times of shut-downs, and in the event of break-downs. The secondary heat disposal system (4 .times. 50%) is located in four vertical shafts or pods distributed symmetrically around the reactor. It is designed to afford a possible by-pass of the main circuit in the event of a break-down, without requiring additional safety valves for a shut off. All valves needed for shutting down the nuclear reactor are advantageously located inside the concrete housing, also placed in vertical pods or shafts, further adding to the safety and compactness of the plant, with these valves being readily accessible from the outside. All components carrying active gas are advantageously integrated inside a safety housing and are accessible while the plant is in operation. The housing has openings required for the installation and removal of components in need of maintenance or repair. The housing, made of pre-stressed concrete, is located in the central area of the safety tank. On the top of this tank is placed a revolving crane used for moving major components in or out of the tank. In the case of a power plant having an output of 1,000 MW, the safety tank is equipped with a cylindrical compartment, which can be sealed off, gas- and pressure-tight, by a simple lid, for holding the generator which is rigidly coupled with the gas turbine assembly. The generator, together with its mounting, can be slid into the compartment, and if necessary, removed again. Further objects, features and advantages of the invention will become apparent from the following detailed description of a preferred embodiment when considered with the attached figures of drawing.
040455262	abstract	A process for the preparation of graphite-clad nuclear fuel rods comprising coating fine particles of a nuclear fuel such as uranium oxide or thorium oxide with a matrix material containing graphite powder and a binder to form matrix material-coated globules, charging the thus-formed globules, without or after being pre-molded into a green rod-like fuel compact, into the rubber mold of a rubber press molding machine, charging a similar matrix material in the gap between the globules or rod-like fuel compact and the inner wall of the rubber mold, compressing the whole mass in the direction from the side face of the whole mass to the axis thereof to form a green, graphite coat material-coated fuel compact, baking the thus-formed green fuel compact at a temperature of up to 1800.degree. C and, if desired, impregnating the baked product with a resin followed by being rebaked at a temperature of up to 1800.degree. C; and the graphite-clad nuclear fuel rods.
	claims	1. A method for operating a feedback control loop and generating a normalized performance index for the feedback control loop, the method comprising:generating, by a feedback controller, an input for a control process comprising building equipment and one or more measurement devices;operating the building equipment to affect a variable state or condition within a building;operating the one or more measurement devices to measure a value of the variable state or condition within the building and generate a feedback signal representing the measured value;identifying, by the feedback controller, an error signal representing a difference between a setpoint and the feedback signal from the control process;computing, by the feedback controller, a first exponentially-weighted moving average (EWMA) of a first function of the error signal and a second EWMA of a second function of the error signal; andgenerating, by the feedback controller, a normalized performance index using the first EWMA and the second EWMA. 2. The method of claim 1, wherein generating the normalized performance index comprises:using the first EWMA to calculate a numerator;using the second EWMA to calculate a denominator; anddividing the numerator by the denominator to generate a normalized ratio of the first EWMA to the second EWMA. 3. The method of claim 1, wherein:the first function of the error signal is the error signal without modification; andthe second function of the error signal is an absolute value of the error signal. 4. The method of claim 3, wherein generating the normalized performance index comprises:calculating an absolute value of the first EWMA; anddividing the absolute value of the first EWMA by the second EWMA. 5. The method of claim 1, wherein:the first function of the error signal is a time-differenced error signal modified by a parameter; andthe second function of the error signal is an absolute value of the error signal. 6. The method of claim 5, further comprising:multiplying the parameter by a value of the error signal from a previous time step; andsubtracting a result of the multiplying from a value of the error signal for a current time step to generate the time-differenced error signal modified by the parameter. 7. The method of claim 5, wherein generating the normalized performance index comprises:using the first EWMA to calculate a numerator;using the second EWMA and the parameter to calculate a denominator; anddividing the numerator by the denominator to generate a normalized ratio of the first EWMA to the second EWMA modified by the parameter. 8. The method of claim 5, further comprising calculating the parameter, wherein calculating the parameter comprises:identifying a sample period and a specified closed loop time constant for the feedback control loop; anddividing the sample period by the specified closed loop time constant. 9. The method of claim 8, further comprising:identifying an integral time parameter for the feedback controller; andusing the integral time parameter as the specified closed loop time constant. 10. The method of claim 8, further comprising:using the error signal to estimate a dominant time constant for the control process; andusing the estimated time constant for the control process as the specified closed loop time constant. 11. The method of claim 1, further comprising:recursively updating the first EWMA and the second EWMA in response to receiving a new measurement of the feedback signal from the control process. 12. A system for operating a feedback control loop and generating a normalized performance index for the feedback control loop, the system comprising:a control process comprising:building equipment operable to affect a variable state or condition within a building; andone or more measurement devices operable to measure a value of the variable state or condition within the building and generate a feedback signal representing the measured value; anda feedback controller having a processing circuit comprising a processor and memory, wherein the processing circuit is configured to:generate an input for the control process;identify an error signal representing a difference between a setpoint and the feedback signal from the control process;compute a first exponentially-weighted moving average (EWMA) of a first function of the error signal and a second EWMA of a second function of the error signal; andgenerate a normalized performance index using the first EWMA and the second EWMA. 13. The system of claim 12, wherein generating the normalized performance index comprises:using the first EWMA to calculate a numerator;using the second EWMA to calculate a denominator; anddividing the numerator by the denominator to generate a normalized ratio of the first EWMA to the second EWMA. 14. The system of claim 12, wherein:the first function of the error signal is the error signal without modification; andthe second function of the error signal is an absolute value of the error signal. 15. The system of claim 14, wherein generating the normalized performance index comprises:calculating an absolute value of the first EWMA; anddividing the absolute value of the first EWMA by the second EWMA. 16. The system of claim 12, wherein:the first function of the error signal is a time-differenced error signal modified by a parameter; andthe second function of the error signal is an absolute value of the error signal. 17. The system of claim 16, wherein generating the normalized performance index comprises:using the first EWMA to calculate a numerator;using the second EWMA and the parameter to calculate a denominator; anddividing the numerator by the denominator to generate a normalized ratio of the first EWMA to the second EWMA modified by the parameter. 18. A system for operating a feedback control loop and generating a normalized performance index for the feedback control loop, the system comprising:a control process comprising:building equipment operable to affect a variable state or condition within a building; andone or more measurement devices operable to measure a value of the variable state or condition within the building and generate a feedback signal representing the measured value; anda feedback controller having a processing circuit comprising a processor and memory, wherein the processing circuit is configured to:identify a first exponentially-weighted moving average (EWMA) and a second EWMA, wherein the first EWMA and the second EWMA are functions of the feedback signal;calculate a ratio of the first EWMA to the second EWMA; andgenerate a normalized performance index using the calculated ratio. 19. The system of claim 18, wherein the processing circuit is configured to:identify an error signal representing a difference between a setpoint and the feedback signal from the control process;calculate the first EWMA using a first function of the error signal; andcalculate the second EWMA using a second function of the error signal. 20. The system of claim 18, wherein the processing circuit is configured to:identify a specified closed loop time constant for the feedback control loop; anduse the specified closed loop time constant to calculate at least one of the first EWMA and the second EWMA.
054948630	claims	1. A process for nuclear waste disposal comprising: providing a glass forming mixture comprising an aqueous solution of one or more metal alkoxides, alcohol, and 35 wt % up to 85 wt % of solubilized, low level radioactive waste in a true solution and having a pH effective to hydrolyze the one or more metal alkoxides; converting the one or more metal alkoxides in the glass forming mixture to a network of corresponding one or more metal oxides; forming a gel from the glass forming mixture containing the network of one or more metal oxides; drying the gel; and sintering the dried gel under conditions effective to form a densified glass. heating the dried gel to a temperature of 550.degree. C. at a rate of 30.degree. C. /hour and heating at a final sintering temperature of 600.degree. to 700.degree. C. reducing the pH of the low level radioactive waste to a value of less than 1 before said providing a glass forming mixture. aging the gel after said forming a gel at a temperature of 20.degree. to 100.degree. C. for a time period of 1 hour to several weeks. reducing the pH of low level radioactive waste to a value of less than 1 to fully solubilize the waste having a reduced pH is an aqueous solution of one or more metal alkoxides and alcohol to form a glass forming mixture; and forming a glass from the glass forming mixture using a sol-gel procedure. adding one or more other glass forming materials to the glass forming mixture after said incorporating. 2. A process according to claim 1, wherein the one or more metal alkoxides is selected from the group consisting of alkoxides of silicon, aluminum, phosphorus, zirconium, boron, germanium, titanium, and mixtures thereof. 3. A process according to claim 2, wherein the one or more metal alkoxides is a mixture comprising an aluminum alkoxide and a silicon alkoxide. 4. A process according to claim 1, wherein the glass forming mixture further comprises compounds containing elements selected from the group consisting of boron, calcium, and mixtures thereof. 5. A process according to claim 1, wherein the low level radioactive waste contains salts of sodium, cesium, molybdenum, strontium, and mixtures thereof. 6. A process according to claim 1, wherein said sintering is carried out at a temperature below that of crystallization onset. 7. A process according to claim 6, wherein said sintering comprises: 8. A process according to claim 1, wherein said forming a gel is carried out at a temperature of 20.degree. to 100.degree. C. 9. A process according to claim 8, wherein said forming a gel is catalyzed by an acid or by a base. 10. A process according to claim 9, wherein said forming a gel is catalyzed by a base. 11. A process according to claim 10, wherein the glass forming mixture comprises N,N-dimethylformamide. 12. A process according to claim 9, wherein said forming a gel is catalyzed by an acid. 13. A process according to claim 12, wherein the glass forming mixture comprises ethylene glycol. 14. A process according to claim 1, wherein said drying is carried out under hypercritical conditions at a temperature of 290.degree. to 310.degree. C. and at a pressure of 136 to 184 atmospheres to form an aerogel. 15. A process according to claim 1, wherein a xerogel forms while temperature of said drying is slowly increased to that of said sintering at substantially atmospheric pressure. 16. A process according to claim 1 further comprising: 17. A process according to claim 1 further comprising: 18. A process for nuclear waste disposal comprising: 19. A process according to claim 18, wherein the one or more metal alkoxides is selected from the group consisting of alkoxides of silicon, aluminum, phosphorus, zirconium, boron, germanium, titanium, and mixtures thereof. 20. A process according to claim 19, wherein the one or more metal alkoxides comprises a mixture of an aluminum alkoxide and a silicon alkoxide. 21. A process according to claim 18, wherein the glass forming mixture further comprises compounds containing elements selected from the group consisting of boron, calcium, and mixtures thereof. 22. A process according to claim 18, wherein the low level radioactive waste contains salts of sodium, cesium, molybdenum, strontium, and mixtures thereof. 23. A process according to claim 18, wherein said reducing comprises adding one or more organic or inorganic acids to the low level radioactive waste. 24. A process according to claim 18 further comprising:

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