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description | This patent is a nationalization of International Patent Application PCT/EP2015/065687, which was filed Jul. 9, 2015 and titled “Container, method for obtaining same and target assembly for the production of radioisotopes using such a container”, which claims priority to Belgian Patent Application BE 2014/0551, which was filed on Jul. 10, 2014 and titled “Conteneur, son procédé d'obtention, et ensemble de cible pour la production de radio-isotopes utilisant un tel conteneur”, both of which are hereby incorporated by reference in their entireties. The invention relates to a container usable for producing radioisotopes, to a method allowing such a container to be obtained, and to a target assembly including such a container. It is known to produce a radioisotope by irradiating a target containing a precursor of the radioisotope by means of a beam of particles. In particular, 18F is produced by irradiating a target material containing 18O-enriched water with a beam of protons. A particle accelerator, such as a cyclotron or a linac, is used to produce the beam of particles. When the precursor of the radioisotope is a liquid or a gas, the target includes a container including a chamber or cavity that is generally closed by a window that allows the beam to pass without being weakened substantially. This window must therefore be as thin as possible, but must withstand the mechanical and thermal stresses and the radiation to which it is subjected in operation. The power dissipated in the target during the irradiation by a beam of particles is given by the product of the energy of the particles by the current of the beam. This power may be very high. The target is generally cooled aggressively by means such as a flow of water. In the case of use of a cyclotron, the target may be placed outside the cyclotron. This solution facilitates the construction of the target and allows easy access to the latter, especially by the cooling means. However, it requires that the beam be extracted from the accelerator, this presenting many difficulties. The various known extracting means, such as stripping, electrostatic or magnetic deflection and self-extraction each also has known difficulties. Extraction by stripping is relatively easy, but requires negative ions that are less stable during the acceleration, more difficult to produce and that require a higher vacuum. Deflectors in general include a septum and a high-voltage electrode that have the function of separating the last turn of the beam from the preceding turn. When the successive turns are closely spaced or overlap, a fraction of the beam strikes the septum, which heats up, is activated and may be damaged. However, once the beam has been extracted, it may be directed toward the target, and it is possible to control the size, the angle and the position of impact of the beam on the target. Another solution consists in placing the target inside the cyclotron. It is then not necessary to extract the beam. The target is placed in the peripheral region of the median plane of the cyclotron. The beam, which traces almost circular orbits of increasing radii, has a certain width and each turn is separated from the preceding turn by a certain distance. This distance may be small, to the point that the beam forms a sort of continuous sheet in the median plane of the cyclotron. A fraction of the beam or of the sheet, which fraction is located radially towards the exterior, then strikes the target, whereas the fraction of the beam or of the sheet that is located radially toward the interior continues to trace its path through the machine. This technique is widely used and with success in the case of solid targets. Document WO 2013049809 discloses a target assembly for producing radioisotopes for the synthesis of radiopharmaceutical products from a liquid precursor. The target, which is shown in FIG. 1, comprises a container 10 including a chamber 12 able to contain a precursor material of the desired radioisotope. A thin covering sheet 14 made of a material that is permeable to the beam covers the chamber and is secured to the container so as to seal the chamber by means of a front clamping flange 16 and a back clamping flange 18. A channel 24 allows access to the chamber 12 for filling or emptying the precursor material. Other securing methods may be envisioned, such as soldering, welding or brazing. The point O represents the center of the cyclotron and the arrow A a beam of particles tracing a turn or an orbit of smaller radius than the radial position of the target. This beam will continue to trace its path through the cyclotron, and reappear with an increased energy and a larger radius. The arrow B represents a more exterior turn, tangentially striking the covering sheet of the target. Some of this beam does not interact with the precursor contained in the chamber, but with the covering sheet 14, thus losing its energy without producing a useful effect. The arrow C represents an even more exterior turn, which penetrates into the chamber 12 and interacts therein with the precursor of the radioisotope that it contains. It may be seen that there is an optimal orientation for the target assembly, minimizing the fraction of beam lost in the tangential edge of the window 14. This implies a precise and therefore difficultly reproducible adjustment of the orientation of the target during each intervention. The assemblage of this target, in particular of the covering sheet, is tricky and the resulting assembly is fragile. When such a covering sheet must be replaced, a technician must intervene on a piece of equipment that has been activated during the irradiation, this requiring time be spent waiting for the radioactivity to decrease. The chamber for the flow of cooling water 20, which is supplied by the tube 22, is placed in thermal contact with the back portion of the chamber 12. The cooling can therefore only be imperfect. Zeisler et al. (Applied Radiation and Isotopes, vol. 53, 2000, pages 449-453) have constructed a spherical target made of niobium in which the beam of particles strikes a first window, consisting of a sheet of aluminum of 0.3 mm thickness, then a layer of cooling water, of 1.1 mm thickness, and lastly the wall of the container, which has the shape of a sphere. This sphere was obtained by welding two hemispheres, themselves obtained by stamping circular blanks made of niobium, of 0.25 mm thickness. Unlike generally known targets, the container of this target does not contain a thin window for the penetration of the beam. The container must on the one hand mechanically resist the pressures that may be generated during the irradiation, and on the other hand be sufficiently thin to decrease the loss of energy of beam. The spherical shape chosen is that which gives the best resistance to pressure, the stresses being uniformly distributed. However, the thickness required to allow the two tubes and two hemispheres to be welded and formed means that the beam loses a significant portion of its energy as it passes therethrough, this producing heat, and meaning that additional cooling of the zone of penetration of the beam is required. This additional cooling is achieved by a flow of water and hence the aluminum window and the layer of water are required, which in turn cause a loss of energy and the production of heat. Because of the need for additional cooling, this target is not suitable for use as an internal target. This target requires a relatively high proton energy (19 MeV) if a significant amount of 18F is to be produced because the loss of energy of these protons in the cooling system and the wall of the container is about 8 MeV. One aim of the invention is to provide a container able to be used for the production of radioisotopes, a method for obtaining such a container, and a target assembly including such a container, that is reliable, easy to assemble and use, and that has a very good transparency to the beam of particles. The invention is defined by the independent claims. The dependent claims define preferred embodiments of the invention. According to a first aspect of the invention, a container is provided for producing radioisotopes by irradiation of a precursor material. According to the invention, the container consists of a metal jacket of integral construction, the wall of said jacket having a thin fraction, of a thickness comprised between 5 and 100 μm, the rest having a thickness larger than 100 μm. In one preferred embodiment, said jacket has a symmetry of revolution, said thin fraction extending over a fraction of the height of the jacket. The container may include at least one end having a conical shape, the base of the cone being oriented toward the exterior of the container. One end of said jacket may be closed. The thin fraction may have an outside diameter comprised between 4 mm and 100 mm. the container may be at least partially made from at least one metal selected from nickel, titanium, niobium, tantalum and the stainless steels. Alloys such as Havar®, Invar® and Kovar® are also preferred. Alloys having a low thermal expansion coefficient are advantageous in the case of rotating targets. According to a second aspect of the invention, a method is provided for obtaining a container according to the invention, which includes the steps of: providing a matrix; electrodepositing on the matrix a thickness of a metallic material, until a first thickness comprised between 5 μm and 100 μm is obtained; masking a fraction of the surface of said matrix; electrodepositing on the unmasked section until a thickness larger than 100 μm is obtained; removing the matrix. The matrix may advantageously be removed by dissolution. According to a third aspect of the invention, a target assembly is provided for producing radioisotopes, including a container according to the invention, and including a holding tube including at one end a threaded portion, and a ring including a suitable interior thread, the holding tube and the ring being configured to encase the container. When the container has an end of conical shape, the holding tube may then advantageously have a conical end congruent with the end of the container, and the ring may advantageously have a conical end congruent with the end of the container. According to one preferred embodiment of the invention, the holding tube and the container are mounted so as to be able to rotate about an axis and the target assembly includes a motor arranged to make the holding tube and the container rotate. The target assembly may include a cooling tube placed inside the container and arranged to allow a cooling liquid to flow. Preferably, the cooling tube may include, at its lower end, a cooling head, which may have on a portion of its periphery liable to receive the beam, a recess, which gives to the incident beam a longer path in a precursor liquid. The target assembly according to the invention may be used as an internal target in a cyclotron or as an external target. It may also be used as a beam stop. FIG. 1 is a cross-sectional view of a prior-art container, namely that of WO2013049809, and was described above. FIG. 2 is a semi-isometric perspective view of a container 100 according to the invention. This container 100 takes the form of a “thimble”, having a symmetry of revolution about an axis. The upper portion 110 is open and may have a conical shape, the opening of the cone being oriented upward. As explained below, this arrangement is of benefit as regards the assemblage of the container 100 into a target assembly. The top of a first cylindrical portion 120 is connected to the upper portion 110 and its bottom is connected to a thin wall section 130. This thin wall section 130 is connected to a second cylindrical portion 140, that itself is connected to a dome 150 closing the container 100 at the bottom. The thickness of the thin fraction is smaller than or equal to 100 μm and for example 80, 60, 40, 20, 10 or even 5 μm. A smaller thickness gives a better transparency to the beam and therefore a better production yield, but is more fragile. The applicant has determined experimentally that the value of 20 μm is a good compromise between these contradictory requirements. The non-thinned portions, namely the open upper portion 110, the first 120 and second 140 cylindrical portion and the dome 150 are produced with a thickness larger than the thickness of the thin wall fraction 130. For example, when the thin fraction has a thickness of 20 μm, the non-thinned portions may have a thickness larger than or equal to 100 μm, 200 μm or more for example. The various portions of the container 100 connect to one another without sharp angles, such that a better mechanical resistance, especially to pressure, is obtained. The inside diameter may be about 10 mm and the total height 11 mm and the angle of the cone may be 30°. The container 100 shown has a cylindrical shape. However, it is possible, without departing from the scope of the present invention, to produce a container 100 having a more complex shape, with a curvature toward the interior, such as a one-sheet hyperboloid, or a bulging shape, such as a barrel. The container 100 has been shown with an upward-facing opening and a closed bottom side. However, it is possible to imagine, without departing from the scope of the invention, a container 100 having two openings such as shown. A container 100 that may be supplied with target material from above or below and through which a coolant fluid or fluid precursor may be made to flow from top to bottom is then obtained. The obtainment of a container 100 according to the invention, in particular when the thin fraction 130 is very thin, presents many difficulties. The applicant has developed a manufacturing method by virtue of which the shape shown, or other shapes, may be produced easily. This method is based on electroforming: A matrix having the shape of the interior of the container 100 is produced. This matrix may for example be made of aluminum; A metal layer is deposited by electrodeposition on all the exterior surface of the matrix, until the thickness desired for the thin portion has been obtained; A fraction of the height of the matrix is masked by applying an insulating layer, a lacquer or a plastic tape for example; the electrodeposition is continued until the thickness desired for the non-thinned portions has been obtained; the matrix is removed, for example in a caustic solution.The thickness of the deposit is determined by the magnitude of the current and the duration of application thereof. The following metals may be used: nickel, titanium, niobium and tantalum, and alloys may also be obtained such as stainless steel, Havar® (cobalt-based alloy), Invar® or Kovar®. In the case of a rotating target, the point of penetration of the beam into the container is a hotspot that is in continuous motion. This spot is a source of thermal expansion/contraction that may lead to fatigue of the metal. The choice of a material with a low thermal expansion coefficient, such as Invar® and Kovar®, may then be advantageous. It is also possible to deposit different alloys or metals in successive electrodeposition steps so as to obtain a first layer in one material, and one or more other layers in other materials. It is thus possible to choose the constituent material of the thin fraction for its resistance to the beam, or to make the layer making contact with the precursor material from a material having a chemical compatibility with the precursor material. Niobium may advantageously be used for the first layer forming the internal wall of the container i.e. the wall making contact with the precursor material. Specifically, it is known that the use of niobium does not lead to contamination of the produced radioisotope by undesired radioisotopes. The choice of the thickness of the thin portion 130 is an important element of the invention. In the table below, the residual energy that a beam of protons having an energy of 7, 10, 15, 20 and 30 MeV, respectively, has after passage through a nickel sheet of various thicknesses has been indicated. It may be seen that when the sheet has a thickness of 5 μm, the energy loss of the protons is negligible i.e. less than 3% at 7 MeV and less than 0.2% at 30 MeV. In contrast, at 100 μm and low energy, the loss in the sheet is substantial. It is then necessary to make recourse to a higher energy and therefore a more expensive accelerator. It is known that the production yield of 18F from H218O by (p,n) reaction is practically zero when the protons have an energy below 3 MeV. To obtain a yield higher than 60 mCi/μA, it is necessary to use protons of 6 MeV at least. The thickness values indicated in bold in the table below are therefore maximum preferred thicknesses, depending on the energy of the available beam. If a yield even higher than 60 mCi/μA is desired, it is necessary to further decrease the thickness of the thin fraction. NICKELSheetIncident E <MeV>thickness710152030<μm>Transmitted E <MeV>56.849.8714.9119.9229.94106.679.7414.8119.8529.89206.329.4814.6219.7029.78405.598.9514.2419.3929.55604.778.3813.8519.0729.33803.867.8013.4318.7629.101002.757.1613.0118.4428.86200Stopped3.0010.7916.7527.72The choice of a thinner wall, for example of thickness smaller than or equal to 100 μm, allows the production of heat as the beam passes through to be limited. The above table may be used to guide the choice of the thickness when the chosen material is nickel. Other metals, such as niobium, titanium or Havar®, have a slightly higher transparency and will give better results. FIG. 3 is an exploded semi-isometric perspective view of the lower portion of a target assembly according to the invention and shows how the container 100 is arranged in a holding tube 200. The tube has a male threaded portion 220. A ring 300 has a corresponding female threaded portion 310. The ring covers the upper portion 110 of the container 100 and presses it against the lower portion of the holding tube 200. At least the thin wall fraction 130 of the container 100 then emerges from the assembly thus formed. The holding tube 200 and the ring 300 may include flats 210, 320 that then allow an operator to assemble and disassemble the assembly very rapidly by means of two open-ended wrenches. The holding tube 200 and the ring 300 may for example be produced from stainless steel. Other mechanical assembling means may also be used without departing from the scope of the invention, such as quick-release hose clamps. In one preferred embodiment of the invention, the lower portion of the holding tube 200 includes a conical end 230 that is congruent with the conical portion 110 of the container 100, said conical portion itself being congruent with a conical end 330 of the ring 300. In this embodiment, an excellent seal tightness may be obtained without having to make recourse to a seal: the seal tightness is ensured by the metal-to-metal contact. FIG. 4 is a cross-sectional view of the lower portion of a target assembly according to the invention. Apart from the elements described above with reference to FIG. 3, the “pocket” assembly 400 is also shown, this pocket assembly playing the dual role of ensuring the cooling of the precursor material contained in the container and that cools in its turn the container, and of allowing the precursor material to be loaded into or unloaded from the container. A cooling tube 410 that is closed at its lower end may be inserted into the holding tube 200 and end in the container 100. In one exemplary embodiment, the container 100 has an inside diameter of 10 mm and a height of 10 mm and the cooling tube 410 an outside diameter of 8 mm, the irradiation chamber 440 having a useful volume of approximately 350 mm3. An intermediate tube 420, which is open at its lower end 425, and of diameter smaller than that of the cooling tube, is inserted into the latter. It is thus possible to make a cooling liquid such as water flow through the space comprised between this cooling tube 410 and this interior tube 420. The arrows A represent the entrance of the cooling liquid and the arrows B the exit of the cooling liquid. The directions of flow A and B may be inverted. Since the heat transfer area is large and uniformly distributed, this arrangement allows excellent cooling to be obtained. In the case where the target assembly allows the assembly made up of the container 100, the holding tube 200 and the ring 300 to be rotated, the “pocket” assembly 400 remains stationary. The relative movement of these 2 assemblies produces a stirring effect that further improves the cooling by inducing a forced convection. A capillary tube 430 placed axially inside the intermediate tube 420 and sealably passing through the lower end of the cooling tube 410 in order to end in the space comprised between the container 100 and the cooling tube 410 allows the precursor material to be loaded and unloaded as indicated by the two-headed arrow C. The enlarged view shows how the conical portion 110 of the container is clamped between the conical end of the ring 330 and the conical end of the holding tube 230, thus ensuring the seal tightness without using a seal. Independently of whether the target of the invention is used as an internal or external target, it is advantageous to be able to make it rotate. It is possible to either successively give thereto various orientations, for example to rotate it by 10° each time it is used, or preferably, to continuously rotate the container 100 during the irradiation. It is thus possible to ensure that all the periphery of the thin wall fraction is passed through by the beam, thereby ensuring a better distribution of the production of heat over a larger area. Furthermore, in the case of a liquid target, the rotation induces stirring of the precursor material, thereby improving the cooling by convection. FIG. 5 is a perspective view of an axial cross section through the upper portion 500 of a target assembly according to the invention, in one embodiment allowing the container 100 to be made to rotate. The container 100 (not shown in the figure) and the holding tube 200 are arranged in the rotor 570 of an electric motor. The stator 560 is secured to a housing 510 that is fixed. Maintenance and seal-tightness are ensured by a seal-bearing having a fixed portion 540 and a rotating portion 542. This seal-bearing may include ball bearings 550 and 550′. This seal may for example be a magnetic fluid seal such as those sold by Rigaku. The distributing head of the pocket 400 emerges from the upper portion of the target assembly and gives access to the orifices 452, 454 through which the cooling fluid respectively enters and exits. and to 430 through which the precursor material is filled/emptied. There may be two separate entrance and exit tubes. FIGS. 6a and 6b show a cyclotron 700 in which a target assembly according to the invention is placed. The upper portion 500 emerges from the upper face of the cyclotron 700. The holding tube 200 has a length such that the container 701 is located in the median plane of the cyclotron, the thin fraction thereof being exposed to the beam, as shown in the detailed view 6c. When the target assembly of the invention is used as an external target, it may be placed at the end of the beamline and receive the beam radially. It is also possible to produce a container the thin portion of which is located on the base, such as in the containers 907 and 909 shown in FIG. 9, and to orient the beam toward this base, parallelly to the axis of symmetry of the container. Certain radioisotope precursors, such as H218O, are precious and expensive. Moreover, it is sometimes advantageous to be able to synthesize radiochemicals from a concentrated product. It is therefore advantageous to minimize the amount used. To this end, a preferred embodiment of the invention has been designed, in which embodiment (shown in FIGS. 7a and 7b) the volume of the chamber is even smaller. FIG. 7a is a semi-isometric perspective view of the lower end of a cooling head 800 of a pocket of this preferred embodiment. This tube has a face 801 having an optimized profile as discussed below. The entrance/exit orifices 802 of the cooling liquid allow the cooling liquid to be made to flow through the interior of the cooling head 800. In this example, there are two parallel entrance and exit tubes, but there could be only a single one thereof as in the example in FIG. 4. The entrance/exit orifices 803 of the precursor liquid open below the lower end of the cooling head 800 and allow the space comprised between the container and the cooling head 800 to be accessed. Notches or grooves 804 may be provided for the placement of temperature probes, thermocouples for example. FIG. 7b is a top view of a cross section perpendicular to the axis of this cooling head 800 in position in a container 860. As may be seen from this cross section, the cooling head 800 has, on a portion of its periphery, a recess 851, which gives to the incident beam, represented by the arrows F, a longer path 852 in the precursor liquid, although the space between the cooling head 800 and the container 160 is smaller in the places where there is no incident beam. The length of this path is defined so that the beam can deposit all its useful energy in the precursor material. This arrangement has the following advantages: decrease of the necessary volume of precursor; maximization of cooling, due to a minimum thickness of liquid; use of all the useful energy (for example the energy higher than 4 MeV for protons in H218O) of the particles of the beam in the precursor. The thermocouples 805 allow the temperature of the target to be controlled in real time. In the embodiment in which the target is rotated, the container 860 rotates whereas the cooling head 800 remain stationary, thereby promoting the stirring of the precursor liquid and the exchange of heat. In this example, the inside diameter of the container 860 is 10 mm, the outside diameter of the cooling head is 9.5 mm and the useful volume of the chamber is 100 mm3. FIG. 9 shows cross-sectional views of a plurality of embodiments of containers according to the invention. The arrow X represents the direction of the incident beam. The arrow X also indicates the position of the thin wall. The cross sections are limited to the facial segment of the solid bodies so as to facilitate the representation of the thin walls. The container 901, which has symmetry of revolution, is cylindrical and has an upper end of conical shape, is one of the preferred embodiments of the invention. The container 902, which has a symmetry of revolution, has two open ends, both of which are of conical shape. The containers 903 and 904 are similar to the container 901, except that they have an open end with a flat edge and an open end with a cylindrical edge, respectively. The container 905 is similar to the container 901, except that it has a “barrel” shape. The container 906 is similar to the container 901, except that it has a one-sheet-hyperboloid shape. The container 907 is similar to the container 901, except that it has a thin wall in the closed end. It thus allows an axial penetration of the beam. The container 908, in contrast to the other containers shown, does not have symmetry of revolution, but a square or rectangular cross section, the thin wall possibly extending over a portion of two or three faces. This container is also shown in semi-isometric perspective. The container 910 is similar to the container 901, except that it has a larger diameter (for example 50 mm) and a flat bottom. The container 909 is similar to the container 910, except that the thin portion is arranged in a ring on the flat bottom and allows an axial penetration of the beam. This container may advantageously be used in an external target, in which the incident beam is parallel to the axis of rotation, as shown by the arrow X. In case of use as an external target, the targets 901 to 907 may be placed such that the beam penetrates into the target radially. The container 100 according to the invention has the advantage of being of integral construction, i.e. of not requiring assembling means or working, mounting or demounting means. The thin fraction 130 of the container 100 forms as it were a window integrated into the container 100. The target and the container 100 according to the invention may be easily demounted and remounted. The operator may act rapidly and may therefore limit his exposure to radiation. The container of the invention requires little material. It is therefore inexpensive and creates little waste when it must be scrapped. The target assembly according to the invention may if needs be serve as a beam stop, for example during the setup of an accelerator. The present invention has been described with reference to specific embodiments, which have been given purely by way a of illustration and which must not be considered to be limiting. Generally, it will appear obvious to those skilled in the art that the present invention is not limited to the examples illustrated and/or described above. The presence of reference numbers in the drawings must not be considered to be limiting, including when these numbers are indicated in the claims. The use of the verbs “comprise”, “contain”, “include”, or any other variant, and their conjugations, in no way excludes the presence of elements other than those mentioned. The use of the indefinite article “a”, “an” or the definite article “the” to introduce an element does not exclude the presence of a plurality of these elements. The use of the words top/bottom lower/upper is to be understood as being relative to the orientation of the components shown in the drawings. Although the examples described relate to the production of 18F by irradiation by a beam of protons of a target material containing 18O-enriched water, the invention may be applied to other liquid precursors, such as ordinary water H216O, which produces 13N during irradiation with protons, or gaseous precursors, such as 14N2 to obtain 11C. It is also possible to apply the invention to pulverulent precursor materials or to powders in suspension in a liquid and forming slurries. Lastly, the invention is also applicable to the case of a precursor material such as 11B2O3, which produces 11C by (p,n) reaction and forms 11CO2 that may be collected. Other particles such as deuterons and alpha particles may be used. Likewise, the target according to the invention may be used with the chamber of the container at atmospheric pressure, or with the chamber placed under pressure. |
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053965337 | description | DESCRIPTION OF THE PREFERRED EMBODIMENTS The principles of the present invention are particularly useful when incorporated in a primary radiation diaphragm illustrated in FIGS. 1 and 2 for a ray beam 2 emitted by a radiation transmitter or emitter 1. For gating the ray beam 2, the primary radiation diaphragm comprises at least one diaphragm plate, but preferably four diaphragm plates, 3, 4, 5 and 6, wherein two of the plates 3 and 4 form a diaphragm plate pair that are arranged in a plane 8 (FIG. 1) and the plates 5 and 6 form a second pair and are arranged in a second plane 7, with the planes 7 and 8 extending parallel to one another and the pairs being offset 90.degree. relative to one another and oppositely adjustable relative to one another. A single mount 9 is provided for the plate 3, a the single mount 10 is provided for the plate 4, a single mount 11 is proved for the plate 5 and a single mount 12 is provided for the plate 6. The mounts 9 and 10 are engaged in a first guideway 13, while the mounts 11 and 12 are engaged in a second guideway 14. As illustrated, the first and second diaphragm plates 3 and 4 are adjusted along the first guideway 13, which proceeds in the plane of the drawing of FIG. 1, while the third and fourth diaphragm plates 5 and 6 are adjusted along a second guideway 14 that proceeds perpendicularly relative to the plane of the drawing of FIG. 1. It is advantageous when the first guideway 13 comprises a first rail 15 and a second rail 16 that are disposed on the same side of the housing 19 of the primary radiation diaphragm and when the second guideway 14 is composed of a first rail 17 and a second rail 18 that are seated on the same side of the housing 19, but extend at right angles to the side having the guideway 13. When the mounts 9, 10, 11 and 12 of the diaphragm plates 3-6 are respectively connected on the first and second guide rails with the mount 9 having a part 20 which engages both the rails 15 and 16, the mount 10 having a part 21 which engages both the rails 15 and 16, while the mount 11 has a part 22 engaging both the rails 17 and 18 and the mount 12 having a part 23 engaging portions of the rails 17 and 18. The bearing support of the diaphragm plates 3 and 4 by the mounts 9 and 10 on the rails 15 and 16 or the bearing supports for the diaphragm plates 5 and 6 by the mounts 11 and 12 on the rails 17 and 18 are set forth in the exemplary embodiment with reference to the third and fourth diaphragm plates 5 and 6, whose mounts 11 and 12, respectively, as well as those of the first and second rails 17 and 18, are illustrated in FIG. 1. It is shown that every guide rail 17 has a first channel 24, a second channel 25 and a third channel 26, while the guide rail 18 has a first channel 27, a second channel 28 and a third channel 29, with these first, second and third channels being adjacent to one another, wherein the guide part, such as 22 of the mount 11 of the third diaphragm plate 5, will have a pair of spaced apart rollers 30 that are received in the first guide channel 24, a single roller 31 received in the guide channel 27 of the second guide rail which lies parallel to the rollers 30 and in the same plane. Also, the mount 22 will have a pair of spaced apart rollers 32 received in the second channel 25 of the first guide rail 17. Analogous to this, the guide part 23 for the mount 12 of the fourth diaphragm plate 6 has a single roller 33 engaged in the third channel 26 of the first guide rail 17, a pair of rollers 34 disposed in the third channel 29 of the second guide rail 18 and also a pair of roller 35 in the second guide channel 28 of the second guide rail 18. This arrangement allows each of the plates, such as 5 and 6, to be held in parallel to each other in the same plane. The oppositely-directed adjustment of the first and second plates 3 and 4 or the third and fourth plates 5 and 6 is set forth in greater detail in FIG. 1, which shows the arrangement for moving the first plates 3 and 4. However, the movement of the third and fourth plates is the same but at right angles. For oppositely-directed adjustment of the first and second diaphragm plates 3 and 4, the mount 9 of the first diaphragm plate 3 is connected to an upper part or run 38a of a toothed belt 38 by a connector, such as a clamp part 36, and the mount 10 of the second diaphragm plate 4 is connected to a lower part or run 38b of the toothed belt 38 by a connector part 37. The two runs 38a and 38b of the belt are aligned parallel to the planes 7 and 8 and extends between an idler roll 40 and a toothed pinion 39a (FIG. 2) on a shaft of a motor 39. Thus, dependent on the direction of rotation of the shaft of the motor 39, the plates 3 and 4 are either adjusted toward one another or away from one another. As illustrated in FIG. 2, a motor 41 with a pinion 41a, an idler 42 and a toothed belt 43 operate in the same way for moving or adjusting the third and fourth diaphragm plates 5 and 6 either together or oppositely away from each other. Within the framework of the invention, each mount 9, 10, 11 and 12 of the diaphragm plates 3, 4, 5 and 6 can also be mounted on the respective guide parts 21-23 at only a single guide rail when the parts 21-23 and the guide rails 15, 16, 17 are executed so that the diaphragm plates are guided in a defined fashion and are, thus, adjustable parallel to the plane. Of course, the channels of the guide rails can also be executed as webs with corresponding grooves or channels engaging therein for the guidance of the diaphragm plates. Although various minor modifications may be suggested by those versed in the art, it should be understood that I wish to embody within the scope of the patent granted hereon all such modifications as reasonably and properly come within the scope of my contribution to the art. |
abstract | The present invention is related to an irradiation cell for producing a radioisotope of interest through the irradiation of a target material by a particle beam, comprising a metallic insert forming a cavity designed to house the target material and to be closed by an irradiation window, wherein said metallic insert comprises at least two separate metallic parts of different materials, being composed of at least a first part comprising said cavity. |
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description | Referring to FIGS. 1 and 2, a computed tomograph (CT) imaging system 10 is shown as including a gantry 12 representative of a xe2x80x9cthird generationxe2x80x9d CT scanner. Gantry 12 has an x-ray source 14 that projects a beam of x-rays 16 toward a detector array 18 on the opposite side of gantry 12. Detector array 18 is formed by detector elements 20 that together sense the projected x-rays that pass through an object 22, for example a medical patient. Each detector element 20 produces an electrical signal that represents the intensity of an impinging x-ray beam and hence the attenuation of the beam as it passes through patient 22. During a scan to acquire x-ray projection data, gantry 12 and the components mounted thereon rotate about a center of rotation or isocenter 24. Rotation of gantry 12 and the operation of x-ray source 14 are governed by a control mechanism 26 of CT system 10. Control mechanism 26 includes an x-ray controller 28 that provides power and timing signals to x-ray source 14 and a gantry motor controller 30 that controls the rotational speed and position of gantry 12. A data acquisition system (DAS) 32 in control mechanism 26 samples analog data from detector elements 20 and converts the data to digital signals for subsequent processing. An image reconstructor 34 receives sampled and digitized x-ray data from DAS 32 and performs high-speed image reconstruction. The reconstructed image is applied as an input to a computer 36 that stores the image in a mass storage device 38. Computer 36 also receives commands and scanning parameters from an operator via console 40 that has a keyboard. An associated cathode ray tube display 42 allows the operator to observe the reconstructed image and other data from computer 36. The operator supplied commands and parameters are used by computer 36 to provide control signals and information to DAS 32, x-ray controller 28 and gantry motor controller 30. In addition, computer 36 operates a table motor controller 44 that controls a motorized table 46 to position patient 22 in gantry 12. Particularly, table 46 moves portions of patient 22 through gantry opening 48. In one embodiment, and as shown in FIG. 3, x-ray beam 16 emanates from a focal spot 50 of x-ray source 14 (FIG. 2). X-ray beam 16 is collimated by collimator 52, and collimated beam 16 is projected toward detector array 18. Detector array 18 is fabricated in a multi-slice configuration and includes detector element rows 54, 56, 58 and 60 for projection data collection. A plane 86, generally referred to as the xe2x80x9cfan beam planexe2x80x9d, contains the centerline of focal spot 50 and the centerline of beam 16. Fan beam plane 86 is illustrated in FIG. 3 as being aligned with a centerline Do of detector array 18, although fan beam plane 86 will not always be so aligned. Detector element rows 62, 64, 66 and 68 serve as z-position detectors for determining a z-axis position of x-ray beam 16. In one embodiment, detector rows 62, 64, 66, and 68 are rows of detector array 18. Outer rows 62 and 68 are selected to be at least substantially within penumbra 70 of beam 16. Inner rows 64 and 66 are selected to be at least substantially within umbra 72 of beam 16. xe2x80x9cAt least substantially withinxe2x80x9d means either entirely within or at least sufficiently within so that outer row 62 and 68 signal intensities depend on an x-ray beam position and inner row 64 and 66 signal intensities provide references against which outer row signals are compared. In one embodiment, collimator 52 includes tapered cams 74 and 76. (Where it is stated herein that a cam xe2x80x9chas a taper,xe2x80x9d it is not intended to exclude cams having a taper of zero unless otherwise stated.) X-ray controller 28 controls positioning of cams 74 and 76. Each cam can be independently positioned to alter position and width of x-ray umbra 72 relative to an edge (not shown) of detector array 18. As shown in FIG. 4, one embodiment of a closed-loop method for positioning beam 16 comprises comparing signals representative of x-ray intensity received from different rows of detector elements and positioning an x-ray beam in accordance with results of the comparison. In one embodiment, signals representative of x-ray intensity from detector rows 62, 64, 66 and 68 are summed 78 to obtain row sums. The summation is over views taken in a 20-millisecond interval. For example, after the analog signals are converted to digital format, hardware circuitry (not shown) in DAS 32 performs offset correction and determines row sums from signals received from outer row 62 and from inner row 64. A corrected ratio R is determined 80 by determining a ratio of a sum of signals received from outer row 62 to a sum of signals received from inner row 64 and multiplying the ratio by a ratio correction factor. The ratio correction factor, determined from imaging system 10 calibration, accounts for different relative DAS gains between outer row 62 and inner row 64. Beam position Z(R) then is determined 82, in millimeters relative to a centerline. Beam position Z is obtained by applying a predetermined beam position transfer function to the corrected ratio to calculate the x-ray beam position. The beam position transfer function Z(R) is represented, for example, by a fourth-degree polynomial having predetermined coefficients: xe2x80x83Z(R)=a+bR+cR2+dR3+eR4 Beam position transfer function Z(R) and its limits are specified at imaging system 10 calibration. A new collimator position is then determined 84. A focal spot position f is determined 84 from beam position Z, current collimator position C and other system geometric parameters in accordance with: f = ( Z - C - T z ) fm zz ( l fs ) + C + T z where Tz represents a current taper of cam 74, fmzz represents a focal spot magnification factor at rows 62 and 64 and is a function of focal spot size, and lfs represents focal spot 50 length. A new position for collimator 52 then is determined 84 for a detector element 20 positioned toward isocenter 24. Collimator 52 is repositioned where an edge (not shown) of collimator 52 would meet a line between focal spot position f and a target beam position Z, which has been specified at imaging system 10 calibration. New collimator position Cn thus is determined in accordance with: C n = ( Z t - f ) cm i ( l fs ) + f xe2x80x83where cmi, represents a current collimator magnification factor at detector element 20 positioned toward isocenter 24 and is a function of focal spot size, and lfs represents focal spot 50 length. In one embodiment, steps 78, 80, 82, and 84 are performed independently for each side of collimator 52 at intervals to continuously obtain new positions for each side of collimator 52. These intervals are, in one embodiment, 20 milliseconds, to sample the x-ray beam 16 position 25 times during a 0.5 second scan to minimize control loop lag error. However, in other embodiments, the interval is between 5 milliseconds and 50 milliseconds. In still other embodiments, the interval is between a minimum value sufficient to avoid effects of quantum noise and high frequency variation (such as due to x-ray tube anode movement at a run frequency between 50 Hz and 160 Hz) and a maximum contrained by a slew rate of the sag curve. Sampling the changing sag curve frequently avoids excessive positioning error. (Sag is a periodic movement of x-ray beam 16 resulting from gravity and from centrifugal forces acting on mechanical structure during a rotation of gantry 12.) During patient scanning, z-position detectors 62, 64, 66 or 68 may become blocked by patient clothing, blankets, or other object. After blockage of a z-position detector 62, 64, 66, or 68 has been detected, or when x-ray source 14 first turns on, the loop sample interval is adjusted downward. In one embodiment, the loop sample interval is adjusted downward to 5 milliseconds. After 4 milliseconds of stabilization, the position of the beam is measured and collimator positioning is started to further minimize initial position errors. During a blockage, loop operation is suspended. To determine if any z-position detectors are blocked, a signal from a last data detector element 90 adjacent a z-position detector 62, 64, 66 or 68 is compared to an expected signal Sx. Z-position detector blockage is assumed, in one embodiment, if a last data detector element 20 signal is less than 0.9 times expected signal Sx. In other embodiments, detector blockage is assumed when a last data detector element 20 signal is less than a value between 0.95 and 0.5 times expected signal Sx. (It is desirable to make this value as large in magnitude as possible to identify patient blockage as quickly as possible, thereby avoiding mispositioning of x-ray beam 16 due to corrupted Z-measurement data. A maximum of 0.95 is used in one embodiment because it is known that x-ray scatter blockage from large patients 22, for example, can reduce a signal to 0.95 times the expected value.) During a blockage, collimator positioning is suspended. However, position measurement continues at an interval that is decreased from 20 to 5 milliseconds. The decreased measurement interval allows imaging system 10 to more quickly detect an end of the blockage and to resume closed-loop positioning. Expected signal Sx is written as: Sx=gmA*csf*t*g, where gmA is a generator current mA signal proportional to an x-ray source 14 energizing current, csf is a scale factor determined at system 10 calibration, t is a DAS sample time period, and g is a gain factor. Gain factor g allows expected signal Sx to be adjusted according to a gain value used for scanning. In one embodiment, this gain value is selectable from a plurality of gain values available in system 10. In one embodiment, closed loop tracking is suspended when signal corruption is detected. Signal corruption is detected, for example, by determining an actual focal spot length from a beam position and a collimator position, and comparing the actual focal spot length to a nominal focal spot length. When a difference of, for example, more than 0.1 millimeter is detected between the actual focal spot length and the nominal spot length, corruption is assumed to exist and collimator positioning is suspended. (In other embodiments, a difference threshold for assuming corruption is as small as 0.05 millimeter or as large as about 0.6 millimeter. In still other embodiments, a value is selected between a lower limit set by higher probabilities of false activation due to noise, x-ray scatter and/or momentary beam position disturbances and an upper limit that still provides some of the advantages of tracking.) However, beam position measurement continues at a decreased interval, as when a blockage is detected. Such corruption may occur, for example, for a short time just prior to or just following detection of a patient blockage. If the corruption persists, for example, over 90xc2x0 of rotation of gantry 12 without detecting a patient blockage, a malfunction of the tracking system requiring servicing has likely occurred. In such an event, a scan is immediately aborted to avoid patient dose and collection of non-diagnostic quality images. In other embodiments, a limit is set from as little as 45xc2x0 to as much as 360xc2x0 of a rotation of gantry 12. In other embodiments, a limit is set between a value at which a false alarm rate due to scatter and/or an occasional exceptionally long partial patient 22 blockage is acceptable and an upper limit representing a design choice as to how long compromised operation (high dose and/or non-diagnostic quality images) can be tolerated before terminating a scan. After system 10 has been switched off, position of focal spot 50 changes as source 14 cools over time. In one embodiment, before system 10 is switched on again, an initial focal spot position is approximated from information obtained when a focal spot position was last measured. An approximation of a linear function is used to model focal spot position change during cooling in one embodiment, and in another embodiment, the linear function is a 97 nanometer per second linear function. Because position change with cooling is an exponential function, the linear approximation is clamped at 0.15 millimeters. This clamping corresponds to approximately 20% of a cooling change in system 10 when fully cold, where a linear approximation to the exponential function suffices. A fully cold position requires 8 to 12 hours without patient scanning, and a tube warm up prior to patient scanning is normally requested if the tube has been off more than 1 hour. Therefore, a fully cold position, although possible, is not likely during normal patient scanning. During tube warm up a current measured position of the focal spot is established again for initial positioning of the collimator. Several tracking loop parameters described herein, specifically, beam position transfer function Z(R) and its limits and target beam position Zt, are determined at system 10 calibration. FIG. 5 illustrates one embodiment of a method for calibrating tracking loop parameters. In this embodiment, data from a stationary sweep scan is collected 100 while collimator 52 is stepped through a sequence of z-axis positions. Beam 16 is incremented 0.3 millimeters on detector array 18 exposure surface for each collimator 52 step position. The sweep scan data is offset-corrected and view averaged 102 to obtain a set of detector samples for each collimator 52 step position. A position of the focal spot is then determined 104. A collimator 52 z-axis position offset from detector array centerline D0 is determined 104, as the point where outer rows 62 and 68 receive signals of half-maximum intensity at full detector element 20 width. Position of focal spot 50 during sweep scan then is determined 104 from collimator 52 z-axis offset and nominal system 10 geometric parameters. A beam 16 position is determined 106 for each detector element 20 at each collimator 52 step position. Beam 16 positions are determined from sweep scan focal spot 50 position, nominal length of focal spot 50, and nominal system 10 geometry. Target beam position Z, then is determined 108 for detector element 20 positioned toward isocenter 24. When beam 16 is directed at target beam position Zt, beam 16 is sufficiently close to detector array 18 edge 92 to prevent imaging artifacts but is far enough away to minimize patient dosage. To determine target beam position Zt, ratios of detector samples for successive collimator 52 step positions are utilized to determine a detector differential error. A reconstruction error sensitivity function w(i) then is applied to weight the detector differential error. Reconstruction error sensitivity function w(i) is related to the percent positive contribution of a detector element 20 as a finction of its radial distance from isocenter 24. Function w(i), in one embodiment, is computed from nominal system geometry. In another embodiment, w(i) is empirically determined. For example, the following equations describe an empirical determination of w(i): b(i)=0.018, 0xe2x89xa6ixe2x89xa65 b(i)=0.035+0.00075x(ixe2x88x925), 5xe2x89xa6ixe2x89xa6213 b(i)=0.414+0.00365x(ixe2x88x92213),214xe2x89xa6ixe2x89xa6n where i represents detector element position from isocenter 24 and b(i) represents an artifact threshold, i.e. a percent differential error, for a double detector element 20 error. Reconstruction error sensitivity function w(i) then is determined in accordance with: xe2x80x83w(i)=0.18/b(i). A collimator 52 step position SP is determined for which the weighted detector differential error exceeds a limit L empirically known to produce image artifacts, for example, 0.04 percent. Target beam position Zt then is set for the isocenter detector element at a distance just preceding SP by an amount exceeding applicable tracking loop positioning error. Beam position transfer function Z(R) then is determined 110 for a ratio R of an average of outer row 62 to inner row 64 signals for a set of detector elements at an extreme end of x-ray fan beam 16. Beam 16 positions, determined xe2x88x92106 for each collimator 52 step position, are fitted to the ratio for each collimator 52 step position with a fourth-degree polynomial, for example, in accordance with: Z(R)=a+bR+cR2+dR3+eR4 over a suitable ratio range between a maximum and minimum for the sequence of steps. A valid position measurement range for Z(R) is determined 112 as between end limits of the set of collimator 52 step positions for which an error between a beam 16 position determined by Z and an actual beam 16 position is less than a predetermined limit, for example, 0.2 millimeters. In other embodiments, the predetermined limit is between 0.1 millimeters to 0.6 millimeters. In still other embodiments, the predetermined limit is set at a value between a lower limit just above a value at which a range of beam 16 position that can be precisely measured is too limited, and just below a lower limit that is deemed to create tracking errors so large as to unacceptably compromise the benefits of tracking. The above-described tracking loop senses the signal ratio between detector rows and moves system collimation to maintain the x-ray beam very close to the imaging system detector array edge during patient scanning. As a result, patient x-ray dosage is reduced 20 to 40 percent without sacrificing image quality. Other functions can be utilized in place of beam position transfer function Z(R) and also in place of reconstruction error sensitivity function w(i). In some embodiments, the methods described herein are implemented by software, firmware, or by a combination thereof controlling either computer 36, image reconstructor 34, or both. Also, additional z-detector rows can be provided. In such an embodiment, various combinations of z-detector row signals can be used as the inner and outer row signals, thereby becoming identified as such, or a different and/or more elaborate transfer function can be used to determine a beam position. The above described calibration methods and apparatus provides improved calibration for z-axis tracking loops for positioning x-ray beams on multi-slice detectors of CT imaging systems. The methods and apparatus provide a target beam position at which to maintain the x-ray beam, a transfer function to convert detector ratio information into a beam position in millimeters (or other suitable units via conversion factors), and valid limits of the ratio to beam position transfer function. It should be understood that system 10 is described herein by way of example only, and the invention can be practiced in connection with other types of imaging systems. Furthermore, it will be recognized by those skilled in the art that the calibration system described herein is also useful for other applications which require x-ray beam tracking calibration, such as for object location or sensing of movement. While the invention has been described in terms of various specific embodiments, those skilled in the art will recognize that the invention can be practiced with modification within the spirit and scope of the claims. |
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051805260 | abstract | A process for cleaning solutions of an alkylphosphate in a hydrophobic organic solvent. The process involves the steps of washing the solution with aqueous sulphuric acid, and contacting the washed solution with an ion exchange material. The ion exchange material may be a resin of the type designated as styrene divinyl benzene copolymers. |
claims | 1. A photolithographic mask for patterning a photosensitive material, comprising:at least one structure region for imaging a structure on the photosensitive material;at least one protective layer made of a chemically and mechanically resistive material disposed on said structure region; andan absorber structure for absorbing incident radiation disposed next to said structure region;said protective layer being made of a thin coating having only a few atomic layers deposited by an atomic layer chemical vapor deposition process so that said protective layer constitutes a negligible alteration of nominal or critical dimensions for the structure region and the absorber structure, and in which additional absorption or reflection losses are negligibly low;said protective layer including at least one material selected from the group consisting of Al2O3, Ta2O5, ZrO2 and HfO2. 2. The photolithographic mask according to claim 1, wherein said protective layer additionally extends over surfaces of said absorber structure. 3. The photolithographic mask according to claim 1, further comprising a carrier element suitable for a transmission exposure of the photosensitive material, said structure region and said absorber structure are disposed on said carrier element. 4. The photolithographic mask according to claim 1, further comprising a carrier element with a reflection device for reflection exposure of the photosensitive material, said structure region and said absorber structure are disposed on said carrier element. 5. The photolithographic mask according to claim 4, wherein said reflection device is a Bragg reflector for electromagnetic radiation having a wavelength of less than 20 nm. 6. The photolithographic mask according to claim 3, wherein said absorber structure has a buffer layer disposed on a surface of said carrier element. 7. The photolithographic mask according to claim 1, wherein said absorber structure contains at least one material selected from the group consisting of Al, Cu, Ti, TiN, Ta, TaN, Ni and Cr. 8. The photolithographic mask according to claim 1, wherein the photosensitive material is disposed on a wafer. 9. The photolithographic mask according to claim 4, wherein said absorber structure has a buffer layer disposed on a surface of said carrier element. 10. A method for fabricating a photolithographic mask for patterning a photosensitive material, which comprises the following steps:providing a carrier element;depositing at least one protective layer made of a chemically and mechanically resistive material using an atomic layer chemical vapor deposition process on a surface of the carrier element, the protective layer being made of a thin coating having only a few atomic layers and including at least one material selected from the group consisting of Al2O3, Ta2O5, ZrO2 and HfO2, so that the protective layer constitutes a negligible alteration of nominal or critical dimensions for an absorber structure, and in which additional absorption or reflection losses are negligibly low; and fabricating the absorber structure, on a surface of the protective layer by deposition and patterning an absorber layer. 11. The method according to claim 10, which further comprises providing the photosensitive material on a wafer. 12. The method according to claim 10, which further comprises depositing an additional protective layer made of a further chemically and mechanically resistive material using the atomic layer deposition process on surfaces of the absorber structure and on the surface of the protective layer. 13. A method for fabricating a photolithographic mask for patterning a photosensitive material, which comprises the following steps:providing a carrier element;depositing at least one protective layer made of a chemically and mechanically resistive material using an atomic layer chemical vapor deposition process on a surface of the carrier element, the protective layer being made of a thin coating having only a few atomic layers and including at least one material selected from the group consisting of Al2O3, Ta2O5, ZrO2 and HfO2, so that the protective layer constitutes a negligible alteration of nominal or critical dimensions for an absorber structure, and in which additional absorption or reflection losses are negligibly low;depositing a buffer layer on a surface of the protective layer;fabricating the absorber structure on a surface of the buffer layer by deposition and patterning an absorber layer;repairing the absorber structure; andremoving the buffer layer in structure regions that are not covered by the absorber structure, by anisotropic etching, the protective layer serving as an etching stop. 14. The method according to claim 13, which further comprises depositing an additional protective layer made of a further chemically and mechanically resistive material using the atomic layer deposition process on surfaces of the absorber structure and on the surface of the protective layer. 15. The method according to claim 13, which further comprises providing the photosensitive material on a wafer. 16. A method for fabricating a photolithographic mask for patterning a photosensitive material, which comprises the following steps:providing a carrier element;fabricating an absorber structure by depositing and patterning an absorber layer on a surface of the carrier element; anddepositing at least one protective layer made of a chemically and mechanically resistive material using an atomic layer chemical vapor deposition process on surfaces of the absorber structure and on surfaces of the carrier element, the protective layer being made of a thin coating having only a few atomic layers and including at least one material selected from the group consisting of Al2O3, Ta2O5, ZrO2 and HfO2, so that the protective layer constitutes a negligible alteration of nominal or critical dimensions for the absorber structure and the carrier element, and in which additional absorption or reflection losses are negligibly low. 17. The method according to claim 16, which further comprises providing the photosensitive material on a wafer. 18. A method for fabricating a photolithographic mask for patterning a photosensitive material, which comprises the following steps:providing a carrier element;depositing a buffer layer on a surface of the carrier element; fabricating an absorber structure on a surface of the buffer layer by depositing and patterning an absorber layer;repairing the absorber structure;removing the buffer layer in structure regions that are not covered by the absorber structure, by anisotropic etching; anddepositing at least one protective layer made of a chemically and mechanically resistive material using an atomic layer chemical vapor deposition process on surfaces of the absorber structure and of the carrier element in the structure regions, the protective layer being made of a thin coating having only a few atomic layers and including at least one material selected from the group consisting of Al2O3, Ta2O5, ZrO2 and HfO2, so that the protective layer constitutes a negligible alteration of nominal or critical dimensions for the absorber structure and the carrier element in the structure regions, and in which additional absorption or reflection losses are negligibly low. 19. The method according to claim 18, which further comprises providing the photosensitive material on a wafer. |
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description | For the purposes of the USPTO extra-statutory requirements, the present application constitutes a continuation-in-part of U.S. patent application Ser. No. 12/228,542, entitled HEAT PIPE NUCLEAR FISSION DEFLAGRATION WAVE REACTOR COOLING, naming Charles E. Ahlfeld, John Rogers Gilleland, Roderick A. Hyde, Muriel Y. Ishikawa, David G. McAlees, Nathan P. Myhrvold, Thomas Allan Weaver, Charles Whitmer and Lowell L. Wood, Jr. as inventors, filed 12 Aug. 2008, now abandoned which is currently co-pending, or is an application of which a currently co-pending application is entitled to the benefit of a filing date. The present application relates to nuclear fission deflagration wave reactor cooling, and systems, applications, apparatuses, and methods related thereto. Illustrative embodiments provide systems, applications, apparatuses, and methods related to nuclear fission deflagration wave reactor cooling. Illustrative embodiments and aspects include, without limitation, nuclear fission deflagration wave reactors, methods of transferring heat of a nuclear fission deflagration wave reactor, methods of transferring heat from a nuclear fission deflagration wave reactor, methods of transferring heat within a nuclear fission deflagration wave reactor, and the like. The foregoing summary is illustrative only and is not intended to be in any way limiting. In addition to the illustrative aspects, embodiments, and features described above, further aspects, embodiments, and features will become apparent by reference to the drawings and the following detailed description. In the following detailed description, reference is made to the accompanying drawings, which form a part hereof. In the drawings, similar symbols typically identify similar components, unless context dictates otherwise. The illustrative embodiments described in the detailed description, drawings, and claims are not meant to be limiting. Other embodiments may be utilized, and other changes may be made, without departing from the spirit or scope of the subject matter presented here. Overview By way of overview, illustrative embodiments provide systems, applications, apparatuses, and methods related to nuclear fission deflagration wave reactor cooling. Illustrative embodiments and aspects include, without limitation, nuclear fission deflagration wave reactors, methods of transferring heat of a nuclear fission deflagration wave reactor, methods of transferring heat from a nuclear fission deflagration wave reactor, methods of transferring heat within a nuclear fission deflagration wave reactor, and the like. Still by way of overview and referring to FIG. 1A, an illustrative nuclear fission deflagration wave reactor 10 will be discussed by way of illustration and not limitation. The illustrative nuclear fission deflagration wave reactor 10 suitably includes a reactor vessel 12. A reactor core assembly 14 is disposed in the reactor vessel 12 and has nuclear fission fuel material disposed therein. At least one primary heat pipe 16 is disposed in thermal communication with the nuclear fission fuel material. Illustrative non-limiting embodiments will now be explained. As a preliminary matter, it will be noted that, for this illustrative example, at least one primary heat pipe 16 is disposed in thermal communication with the nuclear fission fuel material. Thus, in some embodiments, one primary heat pipe 16 may be disposed in thermal communication with the nuclear fission fuel material. Likewise, in some other embodiments, more than one primary heat pipe 16 may be disposed in thermal communication with the nuclear fission fuel material. While the drawings illustrate more than one primary heat pipe 16 included in various embodiments of the nuclear fission deflagration wave reactor 10, such drawings are for illustration purposes only and are not intended to be limiting. To that end, the number of primary heat pipes 16 disposed in thermal communication with the nuclear fission fuel material is not limited in any manner whatsoever. Instead, any number of primary heat pipes 16 may be disposed in thermal communication with the nuclear fission fuel material as desired for a particular application, depending upon without limitation power production requirements, spatial constraints, regulatory restrictions, or the like. For sake of clarity, references to “at least one primary heat pipe 16” in the description that follows (such as in the context of discussions of various embodiments of the nuclear fission deflagration wave reactor 10) will be made to “the primary heat pipes 16”. Nonetheless, it will be appreciated that such references to “the primary heat pipes 16” are made for purposes of clarity and are not intended to limit the number of primary heat pipes 16 to more than one primary heat pipe 16. In some embodiments at least one heat sink 18 may be disposed in thermal communication with the primary heat pipes. 16. The heat sink 18 may be a steam generator, a biomass reactor, or any other processing device that transfers heat from the primary heat pipes 16, as desired. In the example shown by way of illustration and not limitation when the heat sink 18 is a steam generator, a feedwater inlet 20 supplies feedwater 22 to the heat sink 18. Heat is transferred from the primary heat pipes 16 to the feedwater 22, and the feedwater 22 is transformed in phase from liquid to steam 24. The steam 24 exits the heat sink 18 via a steam outlet 26. In some embodiments, the heat sink 18 may be an external heat sink. That is, the heat sink 18 may be disposed external to the reactor vessel 12. In some other embodiments described below, an internal heat sink (not shown in FIG. 1A) may be disposed internal to the reactor vessel 12. It will be appreciated that any number of the heat sinks 18 may be provided as desired for a particular application. For example, as shown in FIG. 1A some embodiments include one heat sink 18. Referring additionally now to FIG. 1B, some embodiments may include two of the heat sinks 18. For sake of brevity, additionally embodiments in which more than two of the heat sinks 18 are not shown. Nonetheless, it will be appreciated that no limit to the number of heat sinks 18 is intended and no limit should be inferred. The number of heat sinks 18 is not limited and any number of the heat sinks 18 may be used as desired for a particular application, depending upon without limitation power production requirements, spatial constraints, regulatory restrictions, or the like. Therefore, for the same clarity reasons as discussed above for the primary heat pipes 16, references will be made to the heat sinks 18 without intention to limit the number of heat sinks 18 to more than one heat sink 18. An overview now has been set forth for some embodiments of the nuclear fission deflagration reactor 10. Next, considerations and an overview will be given by way of example (and not of limitation) regarding a nuclear fission deflagration wave and the nucleonics thereof. Then, additional illustrative details will be given regarding other embodiments and aspects of nuclear fission deflagration wave reactors. Considerations Behind Nuclear Fission Deflagration Wave Reactor Embodiments Before discussing details of the nuclear fission deflagration wave reactor 10, some considerations behind embodiments of the nuclear fission deflagration wave reactor 10 will be given by way of overview but are not to be interpreted as limitations. Some embodiments of the nuclear fission deflagration wave reactor 10 address many of the considerations discussed below. On the other hand, some other embodiments of the nuclear fission deflagration wave reactor 10 may address one, or a select few of these considerations, and need not accommodate all of the considerations discussed below. Portions of the following discussion include information excerpted from a paper entitled “Completely Automated Nuclear Power Reactors For Long-Term Operation: III. Enabling Technology For Large-Scale, Low-Risk, Affordable Nuclear Electricity” by Edward Teller, Muriel Ishikawa, Lowell Wood, Roderick Hyde, and John Nuckolls, presented at the July 2003 Workshop of the Aspen Global Change Institute, University of California Lawrence Livermore National Laboratory publication UCRL-JRNL-122708 (2003) (This paper was prepared for submittal to Energy, The International Journal, 30 Nov. 2003), the contents of which are hereby incorporated by reference. Certain of the nuclear fission fuels envisioned for use in embodiments of the nuclear fission deflagration wave reactor 10 are typically widely available, such as without limitation uranium (natural, depleted, or enriched), thorium, plutonium, or even previously-burned nuclear fission fuel assemblies. Other, less widely available nuclear fission fuels, such as without limitation other actinide elements or isotopes thereof may be used in embodiments of the nuclear fission deflagration wave reactor 10. While some embodiments of the nuclear fission deflagration wave reactor 10 contemplate long-term operation at full power on the order of around ⅓ century to around ½ century or longer, an aspect of some embodiments of the nuclear fission deflagration wave reactor 10 does not contemplate nuclear refueling (but instead contemplate burial in-place at end-of-life) while some aspects of embodiments of the nuclear fission deflagration wave reactor 10 contemplate nuclear refueling—with some nuclear refueling occurring during shutdown and some nuclear refueling occurring during operation at power. It is also contemplated that nuclear fission fuel reprocessing may be avoided in some cases, thereby mitigating possibilities for diversion to military uses and other issues. Other considerations that may affect choices for some embodiments of nuclear fission deflagration wave reactor 10 include disposing in a safe manner long-lived radioactivity generated in the course of operation. It is envisioned that the nuclear fission deflagration wave reactor 10 may be able to mitigate damage due to operator error, casualties such as a loss of coolant accident (LOCA), or the like. In some aspects decommissioning may be effected in low-risk and inexpensive manner. For example, some embodiments of the nuclear fission deflagration wave reactor 10 may entail underground siting, thereby addressing large, abrupt releases and small, steady-state releases of radioactivity into the biosphere. Some embodiments of the nuclear fission deflagration wave reactor 10 may entail minimizing operator controls, thereby automating those embodiments as much as practicable. In some embodiments, a life-cycle-oriented design is contemplated, wherein those embodiments of the nuclear fission deflagration wave reactor 10 can operate from startup to shutdown at end-of-life. In some life-cycle oriented designs, the embodiments may operate in a substantially fully-automatic manner. Embodiments of the nuclear fission deflagration wave reactor 10 lend themselves to modularized construction. Finally, some embodiments of the nuclear fission deflagration wave reactor 10 may be designed according to high power density. Some features of various embodiments of the nuclear fission deflagration wave reactor 10 result from some of the above considerations. For example, simultaneously accommodating desires to achieve ⅓-½ century (or longer) of operations at full power without nuclear refueling and to avoid nuclear fission fuel reprocessing may entail use of a fast neutron spectrum. As another example, in some embodiments a negative temperature coefficient of reactivity (αT) is engineered-in to the nuclear fission deflagration wave reactor 10, such as via negative feedback on local reactivity implemented with strong absorbers of fast neutrons. As a further example, in some embodiments of the nuclear fission deflagration wave reactor 10 a distributed thermostat enables a propagating nuclear fission deflagration wave mode of nuclear fission fuel burn. This mode simultaneously permits a high average burn-up of non-enriched actinide fuels, such as natural uranium or thorium, and use of a comparatively small “nuclear fission igniter” region of moderate isotopic enrichment of nuclear fissionable materials in the core's fuel charge. As another example, in some embodiments of the nuclear fission deflagration wave reactor 10, multiple redundancy is provided in primary and secondary core cooling. Overview of Illustrative Core Nucleonics An overview of (i) the reactor core assembly 14 and its nucleonics and (ii) propagation of a nuclear fission deflagration wave now will be set forth. Given by way of overview and in general terms, structural components of the reactor core assembly 14 may be made of tantalum (Ta), tungsten (W), rhenium (Re), or carbon composite, ceramics, or the like. These materials or similar may be selected to address the high temperatures at which the reactor core assembly 14 typically operates. Alternatively, or additionally, such material selection may be influenced by the materials' creep resistance over the envisioned lifetime of full power operation, mechanical workability, and/or corrosion resistance. Structural components can be made from single materials, or from combinations of materials (e.g., coatings, alloys, multilayers, composites, and the like). In some embodiments, the reactor core assembly 14 operates at sufficiently lower temperatures so that other materials, such as aluminum (Al), steel, titanium (Ti) or the like can be used, alone or in combinations, for structural components. The reactor core assembly 14 suitably can include a nuclear fission igniter and a larger nuclear fission deflagration burn-wave-propagating region. The nuclear fission deflagration burn-wave-propagating region suitably contains thorium or uranium fuel, and functions on the general principle of fast neutron spectrum fission breeding. In some embodiments, uniform temperature throughout the reactor core assembly 14 is maintained by thermostating modules which regulate local neutron flux and thereby control local power production. The reactor core assembly 14 suitably is a breeder for reasons of efficient nuclear fission fuel utilization and of minimization of requirements for isotopic enrichment. Further, and referring now to FIGS. 2A and 2B, the reactor core assembly 14 suitably utilizes a fast neutron spectrum because the high absorption cross-section of fission products for thermal neutrons typically does not permit utilization of more than about 1% of thorium or of the more abundant uranium isotope, 238U, in uranium-fueled embodiments, without removal of fission products. In FIG. 2A, cross-sections for the dominant neutron-driven nuclear reactions of interest for the 232Th-fueled embodiments are plotted over the neutron energy range 10−3-107 eV. It can be seen that losses to radiative capture on fission product nuclei dominate neutron economies at near-thermal (˜0.1 eV) energies, but are comparatively negligible above the resonance capture region (between ˜3-300 eV). Thus, operating with a fast neutron spectrum when attempting to realize a high-gain fertile-to-fissile breeder can help to preclude fuel recycling (that is, periodic or continuous removal of fission products). The radiative capture cross-sections for fission products shown are those for intermediate-Z nuclei resulting from fast neutron-induced fission that have undergone subsequent beta-decay to negligible extents. Those in the central portions of the burn-waves of embodiments of the reactor core assembly 14 will typically have undergone some decay and thus will have somewhat higher neutron avidity. However, parameter studies have indicated that core fuel-burning results may be insensitive to the precise degree of such decay for some configurations. In FIG. 2B, cross-sections for the dominant neutron-driven nuclear reactions of primary interest for the 232Th-fueled embodiments are plotted over the most interesting portion of the neutron energy range, between >104 and <106.5 eV, in the upper portion of FIG. 2B. The neutron spectrum of embodiments of the reactor core assembly 14 peaks in the ≧105 eV neutron energy region. The lower portion of FIG. 2B contains the ratio of these cross-sections vs. neutron energy to the cross-section for neutron radiative capture on 232Th, the fertile-to-fissile breeding step (as the resulting 233Th swiftly beta-decays to 233Pa, which then relatively slowly beta-decays to 233U, analogously to the 239U-239Np-239Pu beta decay-chain upon neutron capture by 238U). It can be seen that losses to radiative capture on fission products can be comparatively negligible over the neutron energy range of interest, and furthermore that atom-fractions of a few tens of percent of high-performance structural material, such as Ta, will impose tolerable loads on the neutron economy in the reactor core assembly 14. These data also suggest that core-averaged fuel burn-up in excess of 50% can be realizable, and that fission product-to-fissile atom-ratios behind the nuclear fission deflagration wave when reactivity is finally driven negative by fission-product accumulation will be approximately 10:1. Origination and Propagation of Nuclear Fission Deflagration Wave Burnfront An illustrative nuclear fission deflagration wave within the reactor core assembly 14 will now be explained. Propagation of deflagration burning-waves through combustible materials can release power at predictable levels. Moreover, if the material configuration has the requisite time-invariant features, the ensuing power production may be at a steady level. Finally, if deflagration wave propagation-speed may be externally modulated in a practical manner, the energy release-rate and thus power production may be controlled as desired. Sustained nuclear fission deflagration waves are rare in nature, due to disassembly of initial nuclear fission fuel configuration as a hydrodynamic consequence of energy release during the earliest phases of wave propagation, in the absence of some control. However, in embodiments of the reactor core assembly 14 a nuclear fission deflagration wave can be initiated and propagated in a sub-sonic manner in fissionable fuel whose pressure is substantially independent of its temperature, so that its hydrodynamics is substantially ‘clamped’. The nuclear fission deflagration wave's propagation speed within the reactor core assembly 14 can be controlled in a manner conducive to large-scale power generation, such as in an electricity-producing reactor system like embodiments of the nuclear fission deflagration wave reactor 10. Nucleonics of the nuclear fission deflagration wave are explained below. Inducing nuclear fission of selected isotopes of the actinide elements—the fissile ones—by capture of neutrons of any energy permits the release of nuclear binding energy at any material temperature, including arbitrarily low ones. The neutrons that are captured by the fissile actinide element may be provided by the nuclear fission igniter. Release of more than a single neutron per neutron captured, on the average, by nuclear fission of substantially any actinide isotope can provide opportunity for a diverging neutron-mediated nuclear-fission chain reaction in such materials. Release of more than two neutrons for every neutron which is captured (over certain neutron-energy ranges, on the average) by nuclear fission by some actinide isotopes may permit first converting an atom of a non-fissile isotope to a fissile one (via neutron capture and subsequent beta-decay) by an initial neutron capture, and then of neutron-fissioning the nucleus of the newly-created fissile isotope in the course of a second neutron capture. Most really high-Z (Z≧90) nuclear species can be combusted if, on the average, one neutron from a given nuclear fission event can be radiatively captured on a non-fissile-but-‘fertile’ nucleus which will then convert (such as via beta-decay) into a fissile nucleus and a second neutron from the same fission event can be captured on a fissile nucleus and, thereby, induce fission. In particular, if either of these arrangements is steady-state, then sufficient conditions for propagating a nuclear fission deflagration wave in the given material can be satisfied. Due to beta-decay in the process of converting a fertile nucleus to a fissile nucleus, the characteristic speed of wave advance is of the order of the ratio of the distance traveled by a neutron from its fission-birth to its radiative capture on a fertile nucleus (that is, a mean free path) to the half-life of the (longest-lived nucleus in the chain of) beta-decay leading from the fertile nucleus to the fissile one. Such a characteristic fission neutron-transport distance in normal-density actinides is approximately 10 cm and the beta-decay half-life is 105-106 seconds for most cases of interest. Accordingly for some designs, the characteristic wave-speed is 10−4-10−7 cm sec−1, or approximately 10−13-10−14 of that of a typical nuclear detonation wave. Such a relatively slow speed-of-advance indicates that the wave can be characterized as a deflagration wave, rather than a detonation wave. If the deflagration wave attempts to accelerate, its leading-edge counters ever-more-pure fertile material (which is relatively lossy in a neutronic sense), for the concentration of fissile nuclei well ahead of the center of the wave becomes exponentially low. Thus the wave's leading-edge (referred to herein as a “burnfront”) stalls or slows. Conversely, if the wave slows, the local concentration of fissile nuclei arising from continuing beta-decay increases, the local rates of fission and neutron production rise, and the wave's leading-edge, that is the burnfront, accelerates. Finally, if the heat associated with nuclear fission is removed sufficiently rapidly from all portions of the configuration of initially fertile matter in which the wave is propagating, the propagation may take place at an arbitrarily low material temperature—although the temperatures of both the neutrons and the fissioning nuclei may be around 1 MeV. Such conditions for initiating and propagating a nuclear fission deflagration wave can be realized with readily available materials. While fissile isotopes of actinide elements are rare terrestrially, both absolutely and relative to fertile isotopes of these elements, fissile isotopes can be concentrated, enriched and synthesized. The use of both naturally-occurring and man-made ones, such as 235U and 239Pu, respectively, in initiating and propagating nuclear fission detonation waves is well-known. Consideration of pertinent neutron cross-sections (shown in FIGS. 2A and 2B) suggests that a nuclear fission deflagration wave can burn a large fraction of a core of naturally-occurring actinides, such as 232Th or 238U, if the neutron spectrum in the wave is a ‘hard’ or ‘fast’ one. That is, if the neutrons which carry the chain reaction in the wave have energies which are not very small compared to the approximately 1 MeV at which they are evaporated from nascent fission fragments, then relatively large losses to the spacetime-local neutron economy can be avoided when the local mass-fraction of fission products becomes comparable to that of the fertile material (recalling that a single mole of fissile material fission-converts to two moles of fission-product nuclei). Even neutronic losses to typical neutron-reactor structural materials, such as Ta, which has desirable high-temperature properties, may become substantial at neutron energies ≦0.1 MeV. Another consideration is the (comparatively small) variation with incident neutron energy of the neutron multiplicity of fission, ν, and the fraction of all neutron capture events which result in fission (rather than merely γ-ray emission). The algebraic sign of the function α(ν-2) constitutes a condition for the feasibility of nuclear fission deflagration wave propagation in fertile material compared with the overall fissile isotopic mass budget, in the absence of neutron leakage from the core or parasitic absorptions (such as on fission products) within its body, for each of the fissile isotopes of the reactor core assembly 14. The algebraic sign is generally positive for all fissile isotopes of interest, from fission neutron-energies of approximately 1 MeV down into the resonance capture region. The quantity α(ν-2)/ν upper-bounds the fraction of total fission-born neutrons which may be lost to leakage, parasitic absorption, or geometric divergence during deflagration wave propagation. It is noted that this fraction is 0.15-0.30 for the major fissile isotopes over the range of neutron energies which prevails in all effectively unmoderated actinide isotopic configurations of practical interest (approximately 0.1-1.5 MeV). In contrast to the situation prevailing for neutrons of (epi-) thermal energy (see FIG. 2B), in which the parasitic losses due to fission products dominate those of fertile-to-fissile conversion by 1-1.5 decimal orders-of-magnitude, fissile element generation by capture on fertile isotopes is favored over fission-product capture by 0.7-1.5 orders-of-magnitude over the neutron energy range 0.1-1.5 MeV. The former suggests that fertile-to-fissile conversion will be feasible only to the extent of 1.5-5% percent at-or-near thermal neutron energies, while the latter indicates that conversions in excess of 50% may be expected for near-fission energy neutron spectra. In considering conditions for propagation of a nuclear fission deflagration wave, in some approaches neutron leakage may be effectively ignored for very large, “self-reflected” actinide configurations. Referring to FIG. 2B and analytic estimates of the extent of neutron moderation-by-scattering entirely on actinide nuclei, it will be appreciated that deflagration wave propagation can be established in sufficiently large configurations of the two types of actinides that are relatively abundant terrestrially: 232Th and 238U, the exclusive and the principal (that is, longest-lived) isotopic components of naturally-occurring thorium and uranium, respectively. Specifically, transport of fission neutrons in these actinide isotopes will likely result in either capture on a fertile isotopic nucleus or fission of a fissile one before neutron energy has decreased significantly below 0.1 MeV (and thereupon becomes susceptible with non-negligible likelihood to capture on a fission-product nucleus). Referring to FIG. 2A, it will be appreciated that fission product nuclei concentrations can significantly exceed fertile ones and fissile nuclear concentrations may be an order-of-magnitude less than the lesser of fission-product or fertile ones while remaining quantitatively substantially reliable. Consideration of pertinent neutron scattering cross-sections suggests that right circular cylindrical configurations of actinides which are sufficiently extensive to be effectively infinitely thick—that is, self-reflecting—to fission neutrons in their radial dimension will have density-radius products >>200 gm/cm2—that is, they will have radii >>10-20 cm of solid-density 238U-232Th. The breeding-and-burning wave provides sufficient excess neutrons to breed new fissile material 1-2 mean-free-paths into the yet-unburned fuel, effectively replacing the fissile fuel burnt in the wave. The ‘ash’ behind the burn-wave's peak is substantially ‘neutronically neutral’, since the neutronic reactivity of its fissile fraction is just balanced by the parasitic absorptions of structure and fission product inventories on top of leakage. If the fissile atom inventory in the wave's center and just in advance of it is time-stationary as the wave propagates, then it is doing so stably; if less, then the wave is ‘dying’, while if more, the wave may be said to be ‘accelerating.’ Thus, a nuclear fission deflagration wave may be propagated and maintained in substantially steady-state conditions for long time intervals in configurations of naturally-occurring actinide isotopes. The above discussion has considered, by way of non-limiting example, circular cylinders of natural uranium or thorium metal of less than a meter or so diameter—and that may be substantially smaller in diameter if efficient neutron reflectors are employed—that may stably propagate nuclear fission deflagration waves for arbitrarily great axial distances. However, propagation of nuclear fission deflagration waves is not to be construed to be limited to circular cylinders, to symmetric geometries, or to singly-connected geometries. To that end, additional embodiments of alternate geometries of nuclear fission deflagration wave reactor cores are described in U.S. patent application Ser. No. 11/605,943, entitled AUTOMATED NUCLEAR POWER REACTOR FOR LONG-TERM OPERATION, naming RODERICK A. HYDE, MURIEL Y. ISHIKAWA, NATHAN P. MYHRVOLD, AND LOWELL L. WOOD, JR. as inventors, filed 28 Nov. 2006, the contents of which are hereby incorporated by reference. Propagation of a nuclear fission deflagration wave has implications for embodiments of the nuclear fission nuclear fission deflagration wave reactor 10. As a first example, local material temperature feedback can be imposed on the local nuclear reaction rate at an acceptable expense in the deflagration wave's neutron economy. Such a large negative temperature coefficient of neutronic reactivity confers an ability to control the speed-of-advance of the deflagration wave. If very little thermal power is extracted from the burning fuel, its temperature rises and the temperature-dependent reactivity falls, and the nuclear fission rate at wave-center becomes correspondingly small and the wave's equation-of-time reflects only a very small axial rate-of-advance. Similarly, if the thermal power removal rate is large, the material temperature decreases and the neutronic reactivity rises, the intra-wave neutron economy becomes relatively undamped, and the wave advances axially relatively rapidly. Details regarding illustrative implementations of temperature feedback that may be incorporated within embodiments of the reactor core assembly 14 are described in U.S. patent application Ser. No. 11/605,933, entitled CONTROLLABLE LONG TERM OPERATION OF A NUCLEAR REACTOR, naming RODERICK A. HYDE, MURIEL Y. ISHIKAWA, NATHAN P. MYHRVOLD, AND LOWELL L. WOOD, JR. as inventors, filed 28 Nov. 2006, the contents of which are hereby incorporated by reference. As a second example of implications of propagation of a nuclear fission deflagration wave on embodiments of the nuclear fission nuclear fission deflagration wave reactor 10, less than all of the total fission neutron production in the nuclear fission nuclear fission deflagration wave reactor 10 may be utilized. For example, the local material-temperature thermostating modules may use around 5-10% of the total fission neutron production in the nuclear fission nuclear fission deflagration wave reactor 10. Another ≦10% of the total fission neutron production in the nuclear fission nuclear fission deflagration wave reactor 10 may be lost to parasitic absorption in the relatively large quantities of high-performance, high temperature, structure materials (such as Ta, W, or Re) employed in structural components of the nuclear fission nuclear fission deflagration wave reactor 10. This loss occurs in order to realize ≧60% thermodynamic efficiency in conversion to electricity and to gain high system safety figures-of-merit. The Zs of these materials, such as Ta, W and Re, are approximately 80% of that of the actinides, and thus their radiative capture cross-sections for high-energy neutrons are not particularly small compared to those of the actinides, as is indicated for Ta in FIGS. 2A and 2B. A final 5-10% of the total fission neutron production in the nuclear fission nuclear fission deflagration wave reactor 10 may be lost to parasitic absorption in fission products. As noted above, the neutron economy characteristically is sufficiently rich that approximately 0.7 of total fission neutron production is sufficient to sustain deflagration wave-propagation in the absence of leakage and rapid geometric divergence. This is in sharp contrast with (epi) thermal-neutron power reactors employing low-enrichment fuel, for which neutron-economy discipline in design and operation must be strict. As a third example of implications of propagation of a nuclear fission deflagration wave on embodiments of the nuclear fission deflagration wave reactor 10, high burn-ups (on the order of around 50% to around 80%) of initial actinide fuel-inventories which are characteristic of the nuclear fission deflagration waves permit high-efficiency utilization of as-mined fuel—moreover without a requirement for reprocessing. Referring now to FIGS. 2C-2G, features of the fuel-charge of embodiments of the reactor core assembly 14 are depicted at four equi-spaced times during the operational life of the reactor after origination of the nuclear fission deflagration wave (referred to herein as “nuclear fission ignition”) in a scenario in which full reactor power is continuously demanded over a ⅓ century time-interval. In the embodiment shown, two nuclear fission deflagration wavefronts propagate from an origination point 28 (near the center of the reactor core assembly 14 and in which the nuclear fission igniter is located) toward ends of the reactor core assembly 14. Corresponding positions of the leading edge of the nuclear fission deflagration wave-pair at various time-points after full ignition of the fuel-charge of the reactor core assembly 14 are indicated in FIG. 2C. FIGS. 2D, 2E, 2F, and 2G illustrate masses (in kg of total mass per cm of axial core-length) of various isotopic components in a set of representative near-axial zones and fuel specific power (in W/g) at the indicated axial position as ordinate-values versus axial position along an illustrative, non-limiting 10-meter-length of the fuel-charge as an abscissal value at approximate times after nuclear fission ignition of approximately 7.5 years, 15 years, 22.5 years, and 30 years, respectively. The central perturbation is due to the presence of the nuclear fission igniter indicated by the origination point 28 (FIG. 2C). It will be noted that the neutron flux from the most intensely burning region behind the burnfront breeds a fissile isotope-rich region at the burnfront's leading-edge, thereby serving to advance the nuclear fission deflagration wave. After the nuclear fission deflagration wave's burnfront has swept over a given mass of fuel, the fissile atom concentration continues to rise for as long as radiative capture of neutrons on available fertile nuclei is considerably more likely than on fission product nuclei, while ongoing fission generates an ever-greater mass of fission products. Nuclear power-production density peaks in this region of the fuel-charge, at any given moment. It will also be noted that in the illustrated embodiments, differing actions of two slightly different types of thermostating units on the left and the right sides of the nuclear fission igniter account for the corresponding slightly differing power production levels. Still referring to FIGS. 2D-2G, it can be seen that well behind the nuclear fission deflagration wave's advancing burnfront, the concentration ratio of fission product nuclei (whose mass closely averages half that of a fissile nucleus) to fissile ones climbs to a value comparable to the ratio of the fissile fission to the fission product radiative capture cross-sections (FIG. 2A), the “local neutronic reactivity” thereupon goes slightly negative, and both burning and breeding effectively cease—as will be appreciated from comparing FIGS. 2D, 2E, 2F, and 2G with each other, far behind the nuclear fission deflagration wave burnfront. In some embodiments of the nuclear fission deflagration wave reactor 10, all the nuclear fission fuel ever used in the reactor is installed during manufacture of the reactor core assembly 14. Also, in some configurations no spent fuel is ever removed from the reactor core assembly 14. In one approach, such embodiments may allow operation without ever accessing the wave reactor core 14 after nuclear fission ignition up to and perhaps after completion of propagation of the burnfront. However, in some other embodiments of the nuclear fission deflagration wave reactor 10, additional nuclear fission fuel may be added to the reactor core assembly 14 after nuclear fission ignition. In some other embodiments of the nuclear fission deflagration wave reactor 10, spent fuel may be removed from the reactor core assembly (and, in some embodiments, removal of spent fuel from the reactor core assembly 14 may be performed while the nuclear fission deflagration wave reactor 10 is operating at power). Such illustrative refueling and defueling is explained in U.S. patent application Ser. No. 11/605,848, entitled METHOD AND SYSTEM FOR PROVIDING FUEL IN A NUCLEAR REACTOR, naming RODERICK A. HYDE, MURIEL Y. ISHIKAWA, NATHAN P. MYHRVOLD, AND LOWELL L. WOOD, JR. as inventors, filed 28 Nov. 2006, the contents of which are hereby incorporated by reference. Regardless of whether or not spent fuel is removed, pre-expansion of the as-loaded fuel permits higher-density actinides to be replaced with lower-density fission products without any overall volume changes in fuel elements, as the nuclear fission deflagration wave sweeps over any given axial element of actinide ‘fuel,’ converting it into fission-product ‘ash.’ Given by way of overview, launching of nuclear fission deflagration waves into 232Th or 238U fuel-charges can initiate with ‘nuclear fission igniter modules’ enriched in fissile isotopes. Illustrative nuclear fission igniter modules and methods for launching nuclear fission deflagration waves are discussed in detail in a co-pending U.S. patent application Ser. No. 12/069,908, entitled NUCLEAR FISSION IGNITER naming CHARLES E. AHLFELD, JOHN ROGERS GILLELAND, RODERICK A. HYDE, MURIEL Y. ISHIKAWA, DAVID G. MCALEES, NATHAN P. MYHRVOLD, CHARLES WITMER, AND LOWELL L. WOOD, JR. as inventors, filed 12 Feb. 2008, the contents of which are hereby incorporated by reference. Higher enrichments can produce more compact modules, and minimum mass modules may employ moderator concentration gradients. In addition, nuclear fission igniter module design may be determined in part by non-technical considerations, such as resistance to materials diversion for military purposes in various scenarios. In other approaches, illustrative nuclear fission igniters may have other types of reactivity sources. For example, other nuclear fission igniters may include “burning embers”, e.g., nuclear fission fuel enriched in fissile isotopes via exposure to neutrons within a propagating nuclear fission deflagration wave reactor. Such “burning embers” may function as nuclear fission igniters, despite the presence of various amounts of fission products “ash”. In other approaches to launching a nuclear fission deflagration wave, nuclear fission igniter modules enriched in fissile isotopes may be used to supplement other neutron sources that use electrically driven sources of high energy ions (such as protons, deuterons, alpha particles, or the like) or electrons that may in turn produce neutrons. In one illustrative approach, a particle accelerator, such as a linear accelerator may be positioned to provide high energy protons to an intermediate material that may in turn provide such neutrons (e.g., through spallation). In another illustrative approach, a particle accelerator, such as a linear accelerator may be positioned to provide high energy electrons to an intermediate material that may in turn provide such neutrons (e.g., by electro-fission and/or photofission of high-Z elements). Alternatively, other known neutron emissive processes and structures, such as electrically induced fusion approaches, may provide neutrons (e.g., 14 Mev neutrons from D-T fusion) that may thereby be used in addition to nuclear fission igniter modules enriched in fissile isotopes to initiate the propagating fission wave. Now that nucleonics of the fuel charge and the nuclear fission deflagration wave have been discussed, further details regarding “nuclear fission ignition” and maintenance of the nuclear fission deflagration wave will be discussed. A centrally-positioned illustrative nuclear fission igniter moderately enriched in fissionable material, such as 235U or 239Pu, has a neutron-absorbing material (such as a borohydride) removed from it (such as by operator-commanded electrical heating), and the nuclear fission igniter becomes neutronically critical. Local fuel temperature rises to a design set-point and is regulated thereafter by the local thermostating modules (discussed in detail in U.S. patent application Ser. No. 11/605,943, entitled AUTOMATED NUCLEAR POWER REACTOR FOR LONG-TERM OPERATION, naming RODERICK A. HYDE, MURIEL Y. ISHIKAWA, NATHAN P. MYHRVOLD, AND LOWELL L. WOOD, JR. as inventors, filed 28 Nov. 2006, the contents of which are hereby incorporated by reference). Neutrons from the fast fission of 235U or 239Pu are mostly captured at first on local 238U or 232Th. It will be appreciated that uranium enrichment of the nuclear fission igniter may be reduced to levels not much greater than that of light water reactor (LWR) fuel by introduction into the nuclear fission igniter and the fuel region immediately surrounding it of a radial density gradient of a refractory moderator, such as graphite. High moderator density enables low-enrichment fuel to burn satisfactorily, while decreasing moderator density permits efficient fissile breeding to occur. Thus, optimum nuclear fission igniter design may involve trade-offs between proliferation robustness and the minimum latency from initial criticality to the availability of full-rated-power from the fully-ignited fuel-charge of the core. Lower nuclear fission igniter enrichments entail more breeding generations and thus impose longer latencies. The peak (unregulated) reactivity of the reactor core assembly 14 slowly decreases in the first phase of the nuclear fission ignition process because, although the total fissile isotope inventory is increasing monotonically, this total inventory is becoming more spatially dispersed. As a result of choice of initial fuel geometry, fuel enrichment versus position, and fuel density, it may be arranged for the maximum reactivity to still be slightly positive at the time-point at which its minimum value is attained. Soon thereafter, the maximum reactivity begins to increase rapidly toward its greatest value, corresponding to the fissile isotope inventory in the region of breeding substantially exceeding that remaining in the nuclear fission igniter. For many cases a quasi-spherical annular shell then provides maximum specific power production. At this point, the fuel-charge of the reactor core assembly 14 can be referred to as “ignited.” Propagation of the nuclear fission deflagration wave, also referred to herein as “nuclear fission burning”, will now be discussed. In the previously described configuration, the spherically-diverging shell of maximum specific nuclear power production continues to advance radially from the nuclear fission igniter toward the outer surface of the fuel charge. When it reaches the outer surface, it typically breaks into two spherical zonal surfaces, with each surface propagating in a respective one of two opposite directions along the axis of the cylinder. At this time-point, the full thermal power production potential of the core may have been developed. This interval is characterized as that of the launching period of the two axially-propagating nuclear fission deflagration wave burnfronts. In some embodiments the center of the core's fuel-charge is ignited, thus generating two oppositely-propagating waves. This arrangement doubles the mass and volume of the core in which power production occurs at any given time, and thus decreases by two-fold the core's peak specific power generation, thereby quantitatively minimizing thermal transport challenges. However, in other embodiments, the core's fuel charge is ignited at or near one end, as desired for a particular application. Such an approach may result in a single propagating wave in some configurations. In other embodiments, the core's fuel charge may be ignited in multiple sites. In yet other embodiments, the core's fuel charge is ignited at any 3-D location within the core as desired for a particular application. In some embodiments, two propagating nuclear fission deflagration waves will be initiated and propagate away from a nuclear fission ignition site, however, depending upon geometry, nuclear fission fuel composition, the action of neutron modifying control structures or other considerations, different numbers (e.g., one, three, or more) of nuclear fission deflagration waves may be initiated and propagated. However, for sake of understanding, the discussion herein refers, without limitation, to propagation of two nuclear fission deflagration wave burnfronts. From this time forward through the break-out of the two waves when they reach or approach the two opposite ends, the physics of nuclear power generation is typically effectively time-stationary in the frame of either wave, as illustrated in FIGS. 2D-2G. The speed of wave advance through the fuel is proportional to the local neutron flux, which in turn is linearly dependent on the thermal power drawn from the reactor core assembly 14 via the collective action on the nuclear fission deflagration wave's neutron budget of the neutron control system, In one approach, the neutron control system may be implemented with thermostating modules (not shown) as has been described in U.S. patent application Ser. No. 11/605,933, entitled CONTROLLABLE LONG TERM OPERATION OF A NUCLEAR REACTOR, naming RODERICK A. HYDE, MURIEL Y. ISHIKAWA, NATHAN P. MYHRVOLD, AND LOWELL L. WOOD, JR. as inventors, filed 28 Nov. 2006, the contents of which are hereby incorporated by reference. When more power is demanded from the reactor via lower-temperature coolant flowing into the core, the temperature of the two ends of the core (which in some embodiments are closest to the coolant inlets) decreases slightly below the thermostating modules' design set-point, a neutron absorber is thereby withdrawn from the corresponding sub-population of the core's thermostating modules, and the local neutron flux is permitted thereby to increase to bring the local thermal power production to the level which drives the local material temperature up to the set-point of the local thermostating modules. However, in the two burnfront embodiment this process is not effective in heating the coolant significantly until its two divided flows move into the two nuclear burn-fronts. These two portions of the core's fuel-charge—which are capable of producing significant levels of nuclear power when not suppressed by the neutron absorbers of the thermostating modules—then act to heat the coolant to the temperature specified by the design set-point of their modules, provided that the nuclear fission fuel temperature does not become excessive (and regardless of the temperature at which the coolant arrived in the core). The two coolant flows then move through the two sections of already-burned fuel centerward of the two burnfronts, removing residual nuclear fission and afterheat thermal power from them, both exiting the fuel-charge at its center. This arrangement encourages the propagation of the two burnfronts toward the two ends of the fuel-charge by “trimming” excess neutrons primarily from the trailing edge of each front, as illustrated in FIGS. 2D-2G. Thus, the core's neutronics in this configuration may be considered to be substantially self-regulated. For example, for cylindrical core embodiments, the core's nucleonics may be considered to be substantially self-regulating when the fuel density-radius product of the cylindrical core is ≧200 gm/cm2 (that is, 1-2 mean free paths for neutron-induced fission in a core of typical composition, for a reasonably fast neutron spectrum). One function of the neutron reflector in such core design may be to substantially reduce the fast neutron fluence seen by the outer portions of the reactor, such as its radiation shield, structural supports, thermostating modules and outermost shell. Theneutron reflector may also impact the performance of the core by increasing the breeding efficiency and the specific power in the outermost portions of the fuel. Such impact may enhance the reactor's economic efficiency. Outlying portions of the fuel-charge are not used at low overall energetic efficiency, but have isotopic burn-up levels comparable to those at the center of the fuel-charge. Final, irreversible negation of the core's neutronic reactivity may be performed at any time by injection of neutronic poison into the coolant stream as desired. For example, lightly loading a coolant stream with a material such as BF3, possibly accompanied by a volatile reducing agent such as H2 if desired, may deposit metallic boron substantially uniformly over the inner walls of coolant-tubes threading through the reactor's core, via exponential acceleration of the otherwise slow chemical reaction 2BF3+3H2->2B+6HF by the high temperatures found therein. Boron, in turn, is a highly refractory metalloid, and will not typically migrate from its site of deposition. Substantially uniform presence of boron in the core in <100 kg quantities may negate the core's neutronic reactivity for indefinitely prolonged intervals without involving the use of powered mechanisms in the vicinity of the reactor. While the core's neutronics in the above-described configurations may be considered to be substantially self-regulated, referring to FIGS. 1C and 1D other configurations may operate under control of a reactor control system 30 that includes a suitable electronic controller 32 having appropriate electrical circuitry and that may include a suitable electro-mechanical system. In a general sense, those skilled in the art will recognize that the various aspects described herein which can be implemented, individually and/or collectively, by a wide range of hardware, software, firmware, and/or any combination thereof can be viewed as being composed of various types of “electrical circuitry.” Consequently, as used herein “electrical circuitry” includes, but is not limited to, electrical circuitry having at least one discrete electrical circuit, electrical circuitry having at least one integrated circuit, electrical circuitry having at least one application specific integrated circuit, electrical circuitry forming a general purpose computing device configured by a computer program (e.g., a general purpose computer configured by a computer program which at least partially carries out processes and/or devices described herein, or a microprocessor configured by a computer program which at least partially carries out processes and/or devices described herein), electrical circuitry forming a memory device (e.g., forms of memory (e.g., random access, flash, read only, etc.)), and/or electrical circuitry forming a communications device (e.g., a modem, communications switch, optical-electrical equipment, etc.). Those having skill in the art will recognize that the subject matter described herein may be implemented in an analog or digital fashion or some combination thereof. In a general sense, those skilled in the art will recognize that the various embodiments described herein can be implemented, individually and/or collectively, by various types of electro-mechanical systems having a wide range of electrical components such as hardware, software, firmware, and/or virtually any combination thereof; and a wide range of components that may impart mechanical force or motion such as rigid bodies, spring or torsional bodies, hydraulics, electro-magnetically actuated devices, and/or virtually any combination thereof. Consequently, as used herein “electro-mechanical system” includes, but is not limited to, electrical circuitry operably coupled with a transducer (e.g., an actuator, a motor, a piezoelectric crystal, a Micro Electro Mechanical System (MEMS), etc.), electrical circuitry having at least one discrete electrical circuit, electrical circuitry having at least one integrated circuit, electrical circuitry having at least one application specific integrated circuit, electrical circuitry forming a general purpose computing device configured by a computer program (e.g., a general purpose computer configured by a computer program which at least partially carries out processes and/or devices described herein, or a microprocessor configured by a computer program which at least partially carries out processes and/or devices described herein), electrical circuitry forming a memory device (e.g., forms of memory (e.g., random access, flash, read only, etc.)), electrical circuitry forming a communications device (e.g., a modem, communications switch, optical-electrical equipment, etc.), and/or any non-electrical analog thereto, such as optical or other analogs. Those skilled in the art will also appreciate that examples of electro-mechanical systems include but are not limited to a variety of consumer electronics systems, medical devices, as well as other systems such as motorized transport systems, factory automation systems, security systems, and/or communication/computing systems. Those skilled in the art will recognize that electro-mechanical as used herein is not necessarily limited to a system that has both electrical and mechanical actuation except as context may dictate otherwise. Illustrative Embodiments of Nuclear Fission Deflagration Wave Reactor Now that some of the considerations behind some of the embodiments of the nuclear fission deflagration wave reactor 10 have been set forth, further details regarding illustrative embodiments of the nuclear fission deflagration wave reactor 10 will be explained. It is emphasized that the following description of illustrative embodiments of the nuclear fission deflagration wave reactor 10 is given by way of non-limiting example only and not by way of limitation. As mentioned above, several embodiments of the nuclear fission deflagration wave reactor 10 are contemplated, as well as further aspects of the nuclear fission deflagration wave reactor 10. After details regarding an illustrative embodiment of the nuclear fission deflagration wave reactor 10 are discussed, other embodiments and aspects will also be discussed. Referring now to FIGS. 1A-1D, the primary heat pipes 16 are disposed in thermal communication with the heat sinks 18. In these arrangements, an evaporator section 34 of the primary heat pipes 16 is disposed in thermal communication with the nuclear fission fuel material (not shown in FIGS. 1A-1D for purposes of clarity). The heat sinks 18 are disposed in thermal communication with a condenser section 36 of the primary heat pipes 16. If desired, the primary heat pipes 16 may also include an adiabatic section 38. Illustrative details of non-limiting aspects of the primary heat pipes 16, such as orientation within the reactor core assembly 14, relationship with the nuclear fission fuel material, and details of illustrative constructions, will be set forth further below. Referring now to FIGS. 3A-3D, in some other embodiments the nuclear fission deflagration wave reactor 10 may also include at least one secondary heat pipe 40 that is disposed in thermal communication with the primary heat pipes 16. In some embodiments at least one heat sink 18 may be disposed in thermal communication with the secondary heat pipes 40. In the examples shown by way of illustration and not limitation when the heat sink 18 is a steam generator, heat is transferred from the secondary heat pipes 40 to the feedwater 22, and the feedwater 22 is transformed in phase from liquid to steam 24. It will be noted that at least one secondary heat pipe 40 is disposed in thermal communication with the primary heat pipes 16. Thus, in some embodiments and similar to the primary heat pipes 16, one secondary heat pipe 40 may be disposed in thermal communication with at least one primary heat pipe 16. Likewise, in some other embodiments, more than one secondary heat pipe 40 may be disposed in thermal communication with the primary heat pipes 16. While the drawings illustrate more than one secondary heat pipe 40 included in various embodiments of the nuclear fission deflagration wave reactor 10, such drawings are for illustration purposes only and are not intended to be limiting. To that end, the number of secondary heat pipes 40 disposed in thermal communication with the primary heat pipes 16 is not limited in any manner whatsoever. Instead, any number of secondary heat pipes 40 may be disposed in thermal communication with the primary heat pipes 16 as desired for a particular application, depending upon without limitation power production requirements, spatial constraints, regulatory restrictions, or the like. For sake of clarity and similar to the primary heat pipes 16, references to “at least one secondary heat pipe 40” in the description that follows (such as in the context of discussions of various embodiments of the nuclear fission deflagration wave reactor 10) will be made to “the secondary heat pipes 40”. Nonetheless, it will be appreciated that such references to “the secondary heat pipes 40” are made for purposes of clarity and are not intended to limit the number of secondary heat pipes 40 to more than one secondary heat pipe 40. As in the examples discussed above, any number of the heat sinks 18 may be provided as desired for a particular application. For example, as shown in FIGS. 3A and 3C some embodiments include one heat sink 18. As shown in FIGS. 3B and 3D, some embodiments may include two of the heat sinks 18. For sake of brevity, additionally embodiments in which more than two of the heat sinks 18 are not shown. Nonetheless, it will be appreciated that no limit to the number of heat sinks 18 is intended and no limit should be inferred. Therefore, for the same clarity reasons as discussed above for the primary heat pipes 16 and the secondary heat pipes 40, references will be made to the heat sinks 18 without intention to limit the number of heat sinks to more than one heat sink 18. While the core's neutronics in the configurations shown in FIGS. 3A and 3B may be considered to be substantially self-regulated, the core's neutronics in the configurations shown in FIGS. 3C and 3D may operate under control of the reactor control system 30 that includes the electronic controller 32 having appropriate electrical circuitry and that may include a suitable electro-mechanical system. These features have been described above, and their details need not be repeated for an understanding thereof. An evaporator section 42 of the secondary heat pipes 40 is disposed in thermal communication with the condenser section 36 of the primary heat pipes 16. The heat sinks 18 are disposed in thermal communication with a condenser section 44 of the secondary heat pipes 40. If desired, the secondary heat pipes 40 may also include an adiabatic section 46. Illustrative details of non-limiting aspects of the secondary heat pipes 40, such as details of illustrative constructions, will be set forth further below. The evaporator section 42 of the secondary heat pipe 40 is disposed in thermal communication with the condenser section 36 of the primary heat pipe 16. That is, heat from the condenser section 36 of the primary heat pipe 16 can be transferred to the evaporator section 42 of the secondary heat pipe 40. Among other things, in order to help maintain physical positioning of the evaporator section 42 of the secondary heat pipe 40 relative to the condenser section 36 of the primary heat pipe 16, in some embodiments the condenser section 36 of the primary heat pipe 16 and the evaporator section 42 of the secondary heat pipe 40 may be disposed within a coupling device 48. In addition to helping maintain physical positioning of the evaporator section 42 of the secondary heat pipe 40 relative to the condenser section 36 of the primary heat pipe 16, the coupling device 48 can also help provide containment in the event of a primary-to-secondary leak. Moreover, the coupling device 48 also can help facilitate transfer of heat from the condenser section 36 of the primary heat pipe 16 to the evaporator section 42 of the secondary heat pipe 40. To that end, the coupling device 48 can help reduce loss of heat to ambient. Further, if desired a heat transfer medium 50 (not shown in FIGS. 3A-3D; see FIGS. 3E-3G and 3I) may be provided within the coupling device 48 to help further facilitate transfer of heat from the condenser section 36 of the primary heat pipe 16 to the evaporator section 42 of the secondary heat pipe 40. Given by way of example and not of limitation, the heat transfer medium 50 may include any heat transfer medium suitable for high temperature operations, such as without limitation 7Li, sodium, potassium, or the like. The condenser section 36 of the primary heat pipe 16 and the evaporator section 42 of the secondary heat pipe 40 may be disposed adjacent each other within the coupling device 48. For example and referring additionally to FIGS. 3E and 3F, in some embodiments the condenser section 36 of the primary heat pipe 16 and the evaporator section 42 of the secondary heat pipe 40 may be disposed laterally adjacent each other within the coupling device 48. As shown in FIG. 3E, the condenser section 36 of the primary heat pipe 16 and the evaporator section 42 of the secondary heat pipe 40 may be disposed laterally adjacent in an end-to-end manner relative to each other. As shown in FIG. 3F, the condenser section 36 of the primary heat pipe 16 and the evaporator section 42 of the secondary heat pipe 40 may be disposed laterally adjacent in an overlapping, “side-to-side” manner relative to each other. In some other embodiments the condenser section 36 of the primary heat pipe 16 and the evaporator section 42 of the secondary heat pipe 40 may be disposed radially adjacent each other within the coupling device 48. Such an arrangement can help provide even further containment in the event of a primary-to-secondary leak. For example and referring additionally to FIGS. 3G and 3H, in some embodiments the condenser section 36 of the primary heat pipe 16 may be radially disposed within the evaporator section 42 of the secondary heat pipe 40. In some other embodiments and referring additionally to FIGS. 3I and 3J, the evaporator section 42 of the secondary heat pipe 40 may be radially disposed within the condenser section 36 of the primary heat pipe 16. Referring now to FIG. 4, in some embodiments one of the heat sinks may be an internal heat sink 52 that is disposed internal to the reactor vessel 12. Thus, the features shown in FIG. 4 can represent a portion of any of the arrangements shown in FIGS. 1A-1D and 3A-3D. The internal heat sink is in thermal communication with an internal heat pipe 54. The internal heat pipe 54 is disposed in thermal communication with the nuclear fission fuel material. As such, the internal heat pipe 54, when provided, may be considered to be one of the primary heat pipes 16. An evaporator section 56 of the internal heat pipe 54 is disposed in thermal communication with the nuclear fission fuel material. The internal heat sink 52 is disposed in thermal communication with a condenser section 58 of the internal heat pipe 54. The internal heat pipe 54 need not include an adiabatic section. In some embodiments, the internal heat pipe 54 includes an adiabatic section (not shown for clarity purposes). In some other embodiments, the internal heat pipe 54 does not include an adiabatic section. The internal heat sink 52 suitably is any type of heat sink as desired for a particular application. In some embodiments the internal heat sink 52 may be a suitable heat transfer device. In some other embodiments the internal heat sink 52 may be a volume of space, which may be at least partially enclosed, within the nuclear reactor vessel 12 in which a workpiece may be placed for heat treatment, annealing, or the like. In some embodiments the internal heat sink 52 may be accessible via an access port 60 defined in the nuclear reactor vessel 12. The primary heat pipes 16 may be arranged in any suitable manner in thermal communication with the nuclear fission fuel material. In general, heat is transferred from the nuclear fission fuel material to the evaporator section 34 of the primary heat pipes 16. Illustrative nuclear fission fuel material and nucleonics of a nuclear fission deflagration wave have been discussed above and need not be repeated. No limitation is to be inferred regarding specific arrangements in which heat is transferred from the nuclear fission fuel material to the primary heat pipes 16. To that end, some illustrative arrangements will be described below and are given by way of non-limiting examples and not by way of limitation. In one arrangement, the primary heat pipes 16 may be disposed external of the nuclear fission fuel material. Referring now to FIGS. 5A -5C by way of non-limiting example, the nuclear fission fuel material may be disposed in nuclear fission fuel assemblies 62. The nuclear fission fuel assemblies 62 may include the nuclear fission fuel material (discussed above), cladding, structural members, and any heat transfer members as desired to facilitate heat transfer from the nuclear fission fuel material toward the primary heat pipes 16. While the nuclear fission fuel assemblies 62 are not shown in FIGS. 1A-1D and 3A-3D for purposes of clarity, in some embodiments the nuclear fission fuel assemblies 62 may be arranged in a matrix of rows and columns. In such an arrangement, the nuclear fission fuel assemblies 62 shown in FIGS. 5A and 5B represent one “slice”—that is, either one row or one column—within the reactor core assembly 14. In such an arrangement, the evaporator section 34 of the primary heat pipes 16 can be arranged substantially perpendicular to the nuclear fission fuel assemblies 62. Thus, in such an arrangement the primary heat pipes 16 also may be arranged in a matrix of rows and columns. The primary heat pipes 16 shown in FIGS. 5A and 5B thus represent one “slice”—that is, one row or one column—within the reactor core assembly 14. A nuclear deflagration wave can be propagated within the reactor core assembly 14 in a manner as described above. In order to help reduce fluence effects, such as without limitation swell, due to slow propagation speed and/or a fast neutron spectrum on components, such as without limitation cladding, of the nuclear fission fuel assemblies 62, it may be desirable for the nuclear fission deflagration wave to propagate perpendicular to (instead of along or parallel to) the nuclear fission fuel assemblies 62. Likewise, it may be desirable for the nuclear fission deflagration wave to propagate perpendicular to (instead of along or parallel to) the primary heat pipes 34 to help reduce any fluence effects on materials or components of the primary heat pipes 16. Thus, in some embodiments the nuclear fission deflagration wave can propagate mutually orthogonal to the nuclear fission fuel assemblies 62 and the primary heat pipes 16. Given by way of non-limiting example and as shown in FIG. 5A, the nuclear fission deflagration wave can propagate into the drawing sheet as indicated by an arrow tail 64. However, as shown in FIG. 5B the nuclear fission deflagration wave can also propagate mutually orthogonal to the nuclear fission fuel assemblies 62 and the primary heat pipes 16 by propagating out of the drawing sheet as indicated by an arrow tip 66. Both directions of the nuclear fission deflagration wave are represented in FIG. 5C. In another illustrative arrangement, the primary heat pipes 16 again are disposed external of the nuclear fission fuel material. Referring now to FIGS. 6A-6C by way of non-limiting example, the nuclear fission fuel material may be disposed in the nuclear fission fuel assemblies 62 as described above. As described above regarding FIGS. 5A-5C, the nuclear fission fuel assemblies 62 may be arranged in a matrix of rows and columns. In such an arrangement, the nuclear fission fuel assemblies 62 shown in FIGS. 6A and 6B represent one “slice”—that is, either one row or one column as illustrated in FIG. 6C—within the reactor core assembly 14. However, in this arrangement, the evaporator section 34 of the primary heat pipes 16 are arranged substantially parallel to the nuclear fission fuel assemblies 62. Thus, in this arrangement the primary heat pipes 16 also may be arranged in a matrix of rows and columns. The primary heat pipes 16 shown in FIGS. 6A-6C thus represent one “slice”—that is, one row or one column—within the reactor core assembly 14. A nuclear deflagration wave can be propagated within the reactor core assembly 14 in a manner as described above. As discussed above, it may be desirable for the nuclear fission deflagration wave to propagate perpendicular to (instead of along or parallel to) the nuclear fission fuel assemblies 62 and the primary heat pipes 34. Thus, in some embodiments the nuclear fission deflagration wave can propagate perpendicular to the nuclear fission fuel assemblies 62 and the primary heat pipes 16. Given by way of non-limiting example and as shown in FIGS. 6A and 6B, the nuclear fission deflagration wave can propagate in either direction away from the arrow tail 64 toward the arrow tip 66. In some arrangements and as shown in FIG. 6D, the primary heat pipes 16 and the nuclear fission fuel assemblies 62 may be located relative to each other such that each nuclear fission fuel assembly 62 is surrounded by primary heat pipes 16. Such an arrangement can help facilitate transfer of heat from the nuclear fission fuel assemblies 62 to the primary heat pipes 16. However, it will be appreciated that the nuclear fission fuel assemblies 62 and the primary heat pipes 16 may be arranged relative to each other in any manner whatsoever as desired for a particular application. Details will now be set forth by way of illustration for several non-limiting examples of primary heat pipes 16, secondary heat pipes 40, and internal heat pipes 54. While several illustrative examples are explained herein, the primary heat pipes 16, secondary heat pipes 40, and internal heat pipes 54 are not to be limited to the illustrative, non-limiting examples described below. Instead, it will be appreciated that any suitable heat pipe may be used as desired for a particular application. The discussion set forth below regarding the illustrative, non-limiting examples of primary heat pipes 16, secondary heat pipes 40, and internal heat pipes 54 is adapted from U.S. patent application Ser. No. 12/152,904, entitled HEAT PIPE FISSION FUEL ELEMENT, naming CHARLES E. AHLFELD, JOHN ROGERS GILLELAND, RODERICK A. HYDE, MURIEL Y. ISHIKAWA, DAVID G. MCALEES, NATHAN P. MYHRVOLD, THOMAS ALLAN WEAVER, CHARLES WHITMER, LOWELL L. WOOD, JR., AND GEORGE B. ZIMMERMAN as inventors, filed 15 May 2008, the contents of which are hereby incorporated by reference. Details of Illustrative Heat Pipes Referring now to FIG. 7A, an illustrative heat pipe can be disposed external of the nuclear fission fuel material. As such, the illustrative heat pipe shown in FIG. 7A may be used as any one or more of the primary heat pipes 16, the secondary heat pipes 40, and/or the internal heat pipes 54. The following discussion explains illustrative details of the non-limiting heat pipe making reference to the primary heat pipes 16, the secondary heat pipes 40, the internal heat pipes 54, and their components. Referring still to FIG. 7A, the heat pipe 16, 40, 54 includes the evaporator section 34, 42, 56 and the condenser section 36, 44, 58. The heat pipe 16, 40, 54 may also include the adiabatic section 38, 46 and (for applications in which the illustrative heat pipe is the internal heat pipe 54) an adiabatic section 68. Heat from the nuclear fission fuel material is transferred to the evaporator section 34, 56 as indicated by arrows 144. Likewise, heat from the condenser section of the primary heat pipes 16 is transferred to the evaporator section 42 as indicated by the arrows 144. The heat pipe 16, 40, 54 defines a cavity 166 therein. A surface 165 of a wall section 163 defines a surface of the cavity 166. The wall section 163 may be made of any suitable material as desired for high-temperature operations and/or, if desired, a neutron flux environment. Given by way of non-limiting example, in some embodiments the wall section 163 may be made of any one or more of materials such as steel, niobium, vanadium, titanium, a refractory metal, and/or a refractory alloy. Given by way of non-limiting example, in some embodiments the refractory metal may be niobium, tantalum, tungsten, hafnium, rhenium, or molybdenum. Non-limiting examples of refractory alloys include, rhenium-tantalum alloys as disclosed in U.S. Pat. No. 6,902,809, tantalum alloy T-111, molybdenum alloy TZM, tungsten alloy MT-185, or niobium alloy Nb-1Zr. A working fluid is provided within the heat pipe 16, 40, 54. The working fluid suitably is evaporable and condensable. Given by way of non-limiting examples, the working fluid may include any suitable working fluid as desired, such as without limitation 7Li, sodium, potassium, or the like. A capillary structure 126 of the heat pipe 16, 40, 54 is defined within at least a portion of the cavity 166. In some embodiments, the capillary structure 126 may be a wick. The wick may be made of any suitable material as desired, such as thorium, molybdenum, tungsten, steel, tantalum, zirconium, carbon, and a refractory metal. In some other embodiments, the capillary structure 126 may be provided as axial grooves. The working fluid in the evaporator section 34, 42, 56 evaporates, as indicated by arrows 146, thereby undergoing phase transformation from a liquid to a gas. The working fluid in gaseous form moves through the heat pipe 16, 40, 54, as indicated by arrows 148, from the evaporator section 34, 42, 56, through the adiabatic section 38, 46, 68, and to the condenser section 36, 44, 58. At the condenser section 36, 44, 58, heat from the working fluid is transferred out of the heat pipe 16, 40, 54, as indicated by arrows 150. The working fluid in the condenser section 36, 44, 58 condenses, as indicated by arrows 152, thereby undergoing phase transformation from a gas to a liquid. The working fluid in liquid form returns from the condenser section 36, 44, 58 through the adiabatic section 38, 46, 68 to the evaporator section 34, 42, 56, as indicated by arrows 154, via capillary action in the capillary structure 126. Referring now to FIG. 7B, in some other embodiments an illustrative heat pipe is similar to that shown in FIG. 7A and described above. However, the heat pipe shown in FIG. 7B does not include an adiabatic section. All other features are similar to those shown in FIG. 7A. To that end, the working fluid in the evaporator section 34, 42, 56 evaporates, as indicated by the arrows 146, thereby undergoing phase transformation from a liquid to a gas. The working fluid in gaseous form moves through the heat pipe 16, 40, 54, as indicated by the arrow 148, from the evaporator section 34, 42, 56 to the condenser section 36, 44, 58. At the condenser section 36, 44, 58, heat from the working fluid is transferred out of the heat pipe 16, 40, 54, as indicated by the arrows 150. The working fluid in the condenser section 36, 44, 58 condenses, as indicated by the arrows 152, thereby undergoing phase transformation from a gas to a liquid. The working fluid in liquid form returns from the condenser section 36, 44, 58 to the evaporator section 34, 42, 56, as indicated by the arrows 154, via capillary action in the capillary structure 126. It will be appreciated that the illustrative heat pipe shown in FIG. 7B can be used as the primary heat pipe 16 or the secondary heat pipe 40, as desired for a particular application. However, it may be desirable to use the illustrative heat pipe shown in FIG. 7B as the internal heat pipe 54 if size constraints are a consideration. Referring now to FIG. 8A, in some other embodiments nuclear fission fuel material 164 may be disposed in at least a portion of a heat pipe. Because the nuclear fission fuel material 164 is disposed in a portion therein, the illustrative heat pipe shown in FIG. 8A may be used as the primary heat pipe 16 or the internal heat pipe 54. The heat pipe 16, 54 defines a cavity 166 therein. The surface 165 of the wall section 163 defines a surface of the cavity 166. In some embodiments, the nuclear fission fuel material 164 is disposed within at least a portion of the cavity 166. For example, in some embodiments the nuclear fission fuel material 164 may be disposed within the capillary structure 126. However, it will be appreciated that the nuclear fission fuel material 164 need not be disposed within the capillary structure 126 and may be disposed anywhere whatsoever within the cavity 166 as desired. In some embodiments, given by way of non-limiting example the nuclear fission fuel material 164 may have a capillary structure. If desired, in some other embodiments the nuclear fission fuel material 164 may have a sintered powdered fuel microstructure, or a foam microstructure, or a high density microstructure, or the like. In some other embodiments a portion of the wall section 163 can include the nuclear fission fuel material 164. In such arrangements the nuclear fission fuel material 164 can be disposed outside of the cavity 166. With the exception of addition of the nuclear fission fuel material 164, other features shown in FIG. 8A are similar to those shown in FIG. 7A. To that end, the working fluid in the evaporator section 34, 56 evaporates, as indicated by the arrows 146, thereby undergoing phase transformation from a liquid to a gas. The working fluid in gaseous form moves through the heat pipe 16, 54, as indicated by the arrows 148, from the evaporator section 34, 56, through the adiabatic section 38, 68, and to the condenser section 36, 58. At the condenser section 36, 58, heat from the working fluid is transferred out of the heat pipe 16, 54, as indicated by the arrows 150. The working fluid in the condenser section 36, 58 condenses, as indicated by the arrows 152, thereby undergoing phase transformation from a gas to a liquid. The working fluid in liquid form returns from the condenser section 36, 58 through the adiabatic section 38, 68 to the evaporator section 34, 56, as indicated by the arrows 154, via capillary action in the capillary structure 126. Referring now to FIG. 8B, in some other embodiments an illustrative heat pipe is similar to that shown in FIG. 8A and described above. However, the heat pipe shown in FIG. 8B does not include an adiabatic section. All other features are similar to those shown in FIG. 8A. To that end, the working fluid in the evaporator section 34, 56 evaporates, as indicated by the arrows 146, thereby undergoing phase transformation from a liquid to a gas. The working fluid in gaseous form moves through the heat pipe 16, 54, as indicated by the arrow 148, from the evaporator section 34, 56 to the condenser section 36, 58. At the condenser section 36, 58, heat from the working fluid is transferred out of the heat pipe 16, 54, as indicated by the arrows 150. The working fluid in the condenser section 36, 58 condenses, as indicated by the arrows 152, thereby undergoing phase transformation from a gas to a liquid. The working fluid in liquid form returns from the condenser section 36, 58 to the evaporator section 34, 56, as indicated by the arrows 154, via capillary action in the capillary structure 126. It will be appreciated that the illustrative heat pipe shown in FIG. 8B can be used as the primary heat pipe 16 as desired for a particular application. However, it may be desirable to use the illustrative heat pipe shown in FIG. 8B as the internal heat pipe 54 if size constraints are a consideration. Referring now to FIGS. 9A and 9B, in some other embodiments at least a portion 214 (shown in phantom) of an illustrative heat pipe may be disposed in a portion of nuclear fission fuel material 212. Because at least the portion 214 of the heat pipe is disposed in a portion of the nuclear fission fuel material 212, the illustrative heat pipe shown in FIG. 9A may be used as the primary heat pipe 16 or the internal heat pipe 54. At least the portion 214 of the heat pipe 16, 54 may be defined by a cavity 218 that may be defined in the nuclear fission fuel material 212. In some embodiments, the cavity 218 may be a passageway that is defined through at least the portion 214 of the nuclear fission fuel material 212. Thus, in some embodiments, a surface 220 of the cavity 218 may be a wall of the portion 214 of the heat pipe 16, 54. The cavity 218 may be defined in any suitable manner. For example, in some embodiments the cavity 218 may be defined by machining the cavity from the nuclear fission fuel material 212 in any manner as desired, such as by drilling, milling, stamping, or the like. In some other embodiments the cavity 218 may be defined by forming at least a portion 222 of the nuclear fission fuel material 212 around a shape, such as without limitation a mandrel (not shown). The forming may be performed in any manner as desired, such as without limitation by welding, casting, electroplating, pressing, molding, or the like. Referring additionally to FIG. 10A, the surface 165 of the wall section 163 of the heat pipe 16, 54 extends from the cavity 218 in the nuclear fission fuel material 212, thereby substantially acting as an extension of the surface 220. As such, the cavity 218 can be considered to be substantially sealed. The capillary structure 126 of the heat pipe 16, 54 is defined within at least a portion of the cavity 218. That is, the surface 220 is a wall that surrounds a portion of the capillary structure 126. In some embodiments, the capillary structure 126 may also be defined in an interior of the heat pipe 16, 54 that is outside the nuclear fission fuel material 212 and enclosed by the wall section 163. In some embodiments, the capillary structure 126 may be a wick. The wick may be made of any suitable material as desired, such as thorium, molybdenum, tungsten, steel, tantalum, zirconium, carbon, and a refractory metal. In some other embodiments, the capillary structure 126 may be provided as axial grooves. A working fluid is provided within the heat pipe 16, 54. The working fluid suitably is evaporable and condensable. Given by way of non-limiting examples, the working fluid may include any suitable working fluid as desired, such as without limitation 7Li, sodium, potassium, or the like. Heat from the nuclear fission fuel material 212 is transferred to the evaporator section 34, 56 as indicated by the arrows 144. The working fluid in the evaporator section 34, 56 evaporates, as indicated by the arrows 146, thereby undergoing phase transformation from a liquid to a gas. The working fluid in gaseous form moves through the heat pipe 16, 54, as indicated by the arrows 148, from the evaporator section 34, 56, through the adiabatic section 38, 68, and to the condenser section 36, 58. At the condenser section 36, 58, heat from the working fluid is transferred out of the heat pipe 16, 54, as indicated by the arrows 150. The working fluid in the condenser section 36, 58 condenses, as indicated by the arrows 152, thereby undergoing phase transformation from a gas to a liquid. The working fluid in liquid form returns from the condenser section 36, 58 through the adiabatic section 38, 68 to the evaporator section 34, 56, as indicated by the arrows 154, via capillary action in the capillary structure 126. Referring now to FIG. 10B, in some other embodiments an illustrative heat pipe is similar to that shown in FIG. 10A and described above. However, the heat pipe shown in FIG. 10B does not include an adiabatic section. All other features are similar to those shown in FIG. 10A. To that end, heat from the nuclear fission fuel material 212 is transferred to the evaporator section 34, 56 as indicated by the arrows 144. The working fluid in the evaporator section 34, 56 evaporates, as indicated by the arrows 146, thereby undergoing phase transformation from a liquid to a gas. The working fluid in gaseous form moves through the heat pipe 16, 54, as indicated by the arrow 148, from the evaporator section 34, 56 to the condenser section 36, 58. At the condenser section 36, 58, heat from the working fluid is transferred out of the heat pipe 16, 54, as indicated by the arrows 150. The working fluid in the condenser section 36, 58 condenses, as indicated by the arrows 152, thereby undergoing phase transformation from a gas to a liquid. The working fluid in liquid form returns from the condenser section 36, 58 to the evaporator section 34, 56, as indicated by the arrows 154, via capillary action in the capillary structure 126. It will be appreciated that the illustrative heat pipe shown in FIG. 10B can be used as the primary heat pipe 16 as desired for a particular application. However, it may be desirable to use the illustrative heat pipe shown in FIG. 10B as the internal heat pipe 54 if size constraints are a consideration. Illustrative Methods Now that illustrative embodiments of nuclear fission deflagration wave reactors and illustrative, non-limiting heat pipes for use therewith have been discussed, illustrative methods associated therewith will now be discussed. Following are a series of flowcharts depicting implementations of processes. For ease of understanding, the flowcharts are organized such that the initial flowcharts present implementations via an overall “big picture” viewpoint and thereafter the following flowcharts present alternate implementations and/or expansions of the “big picture” flowcharts as either sub-steps or additional steps building on one or more earlier-presented flowcharts. Those having skill in the art will appreciate that the style of presentation utilized herein (e.g., beginning with a presentation of a flowchart(s) presenting an overall view and thereafter providing additions to and/or further details in subsequent flowcharts) generally allows for a rapid and easy understanding of the various process implementations. In addition, those skilled in the art will further appreciate that the style of presentation used herein also lends itself well to modular design paradigms. Referring now to FIG. 11 A, an illustrative method 310 is provided for transferring heat of a nuclear fission deflagration wave reactor. The method 310 starts at a block 312. At a block 314 a nuclear fission deflagration wave is propagated in nuclear fission fuel material in a reactor core assembly of a nuclear fission deflagration wave reactor. At a block 316 heat from the nuclear fission fuel material is transferred to at least one primary heat pipe. Given by way of illustration and not of limitation, the heat can be transferred from a portion of the nuclear fission fuel material that is proximate a burnfront of the nuclear fission deflagration wave. The method 310 stops at a block 318. Referring now to FIG. 11B, at a block 320 heat can be transferred from the at least one primary heat pipe to at least one external heat sink that is external of a reactor vessel. Referring now to FIG. 11C, at a block 322 heat can be transferred from the at least one primary heat pipe to at least one secondary heat pipe that is external of a reactor vessel. At a block 324 heat can be transferred from the at least one secondary heat pipe to at least one external heat sink that is external of the reactor vessel. Referring now to FIG. 11D, at a block 326 heat can be transferred from the nuclear fission fuel material to at least one internal heat pipe that is disposed internal to a reactor vessel. At a block 328 heat can be transferred from the at least one internal heat pipe to at least one internal heat sink that is disposed internal to the reactor vessel. Referring now to FIG. 12A, an illustrative method 330 is provided for transferring heat from a nuclear fission deflagration wave reactor. The method 330 starts at a block 332. At a block 334 a nuclear fission deflagration wave is propagated in nuclear fission fuel material in a reactor core assembly of a nuclear fission deflagration wave reactor. At a block 336 heat is transferred from the nuclear fission fuel material to at least one primary heat pipe. Given by way of illustration and not of limitation, the heat can be transferred from a portion of the nuclear fission fuel material that is proximate a burnfront of the nuclear fission deflagration wave. At a block 338 heat is transferred from the at least one primary heat pipe to at least one external heat sink that is external of a reactor vessel. The method 330 stops at a block 340. Referring now to FIG. 12B, at a block 342 heat can be transferred from the at least one primary heat pipe to at least one secondary heat pipe that is external of a reactor vessel. At a block 344 heat is transferred from the at least one secondary heat pipe to at least one external heat sink that is external of the reactor vessel. Referring now to FIG. 12C, at a block 346 heat can be transferred from the nuclear fission fuel material to at least one internal heat pipe that is disposed internal to a reactor vessel. At a block 348 heat is transferred from the at least one internal heat pipe to at least one internal heat sink that is disposed internal to the reactor vessel. Referring now to FIG. 13A, an illustrative method 350 is provided for transferring heat from a nuclear fission deflagration wave reactor. The method 350 starts at a block 352. At a block 354 a nuclear fission deflagration wave is propagated in nuclear fission fuel material in a reactor core assembly of a nuclear fission deflagration wave reactor. At a block 356 heat is transferred from the nuclear fission fuel material to at least one primary heat pipe. Given by way of illustration and not of limitation, the heat can be transferred from a portion of the nuclear fission fuel material that is proximate a burnfront of the nuclear fission deflagration wave. At a block 358 heat is transferred from the at least one primary heat pipe to at least one secondary heat pipe that is external of a reactor vessel. At a block 360 heat is transferred from the at least one secondary heat pipe to at least one external heat sink that is external of the reactor vessel. The method 350 stops at a block 362. Referring now to FIG. 13B, at a block 364 heat can be transferred from the nuclear fission fuel material to at least one internal heat pipe that is disposed internal to a reactor vessel. At a block 366 heat is transferred from the at least one internal heat pipe to at least one internal heat sink that is disposed internal to the reactor vessel. Referring now to FIG. 14, an illustrative method 370 is provided for transferring heat within a nuclear fission deflagration wave reactor. The method 370 begins at a block 372. At a block 374 a nuclear fission deflagration wave is propagated in nuclear fission fuel material in a reactor core assembly of a nuclear fission deflagration wave reactor. At a block 376 heat is transferred from the nuclear fission fuel material to at least one internal heat pipe that is disposed internal to a reactor vessel. Given by way of illustration and not of limitation, the heat can be transferred from a portion of the nuclear fission fuel material that is proximate a burnfront of the nuclear fission deflagration wave. At a block 378 heat is transferred from the at least one internal heat pipe to at least one internal heat sink that is disposed internal to the reactor vessel. The method 370 stops at a block 380. One skilled in the art will recognize that the herein described components (e.g., blocks), devices, and objects and the discussion accompanying them are used as examples for the sake of conceptual clarity and that various configuration modifications are within the skill of those in the art. Consequently, as used herein, the specific exemplars set forth and the accompanying discussion are intended to be representative of their more general classes. In general, use of any specific exemplar herein is also intended to be representative of its class, and the non-inclusion of such specific components (e.g., blocks), devices, and objects herein should not be taken as indicating that limitation is desired. With respect to the use of substantially any plural and/or singular terms herein, those having skill in the art can translate from the plural to the singular and/or from the singular to the plural as is appropriate to the context and/or application. The various singular/plural permutations are not expressly set forth herein for sake of clarity. While particular aspects of the present subject matter described herein have been shown and described, it will be apparent to those skilled in the art that, based upon the teachings herein, changes and modifications may be made without departing from the subject matter described herein and its broader aspects and, therefore, the appended claims are to encompass within their scope all such changes and modifications as are within the true spirit and scope of the subject matter described herein. Furthermore, it is to be understood that the invention is defined by the appended claims. It will be understood by those within the art that, in general, terms used herein, and especially in the appended claims (e.g., bodies of the appended claims) are generally intended as “open” terms (e.g., the term “including” should be interpreted as “including but not limited to,” the term “having” should be interpreted as “having at least,” the term “includes” should be interpreted as “includes but is not limited to,” etc.). It will be further understood by those within the art that if a specific number of an introduced claim recitation is intended, such an intent will be explicitly recited in the claim, and in the absence of such recitation no such intent is present. For example, as an aid to understanding, the following appended claims may contain usage of the introductory phrases “at least one” and “one or more” to introduce claim recitations. However, the use of such phrases should not be construed to imply that the introduction of a claim recitation by the indefinite articles “a” or “an” limits any particular claim containing such introduced claim recitation to inventions containing only one such recitation, even when the same claim includes the introductory phrases “one or more” or “at least one” and indefinite articles such as “a” or “an” (e.g., “a” and/or “an” should typically be interpreted to mean “at least one” or “one or more”); the same holds true for the use of definite articles used to introduce claim recitations. In addition, even if a specific number of an introduced claim recitation is explicitly recited, those skilled in the art will recognize that such recitation should typically be interpreted to mean at least the recited number (e.g., the bare recitation of “two recitations,” without other modifiers, typically means at least two recitations, or two or more recitations). Furthermore, in those instances where a convention analogous to “at least one of A, B, and C, etc.” is used, in general such a construction is intended in the sense one having skill in the art would understand the convention (e.g., “a system having at least one of A, B, and C” would include but not be limited to systems that have A alone, B alone, C alone, A and B together, A and C together, B and C together, and/or A, B, and C together, etc.). In those instances where a convention analogous to “at least one of A, B, or C, etc.” is used, in general such a construction is intended in the sense one having skill in the art would understand the convention (e.g., “a system having at least one of A, B, or C” would include but not be limited to systems that have A alone, B alone, C alone, A and B together, A and C together, B and C together, and/or A, B, and C together, etc.). It will be further understood by those within the art that virtually any disjunctive word and/or phrase presenting two or more alternative terms, whether in the description, claims, or drawings, should be understood to contemplate the possibilities of including one of the terms, either of the terms, or both terms. For example, the phrase “A or B” will be understood to include the possibilities of “A” or “B” or “A and B.” With respect to the appended claims, those skilled in the art will appreciate that recited operations therein may generally be performed in any order. Examples of such alternate orderings may include overlapping, interleaved, interrupted, reordered, incremental, preparatory, supplemental, simultaneous, reverse, or other variant orderings, unless context dictates otherwise. With respect to context, even terms like “responsive to,” “related to,” or other past-tense adjectives are generally not intended to exclude such variants, unless context dictates otherwise. While various aspects and embodiments have been disclosed herein, other aspects and embodiments will be apparent to those skilled in the art. The various aspects and embodiments disclosed herein are for purposes of illustration and are not intended to be limiting, with the true scope and spirit being indicated by the following claims. |
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044951477 | claims | 1. Heat retarding closure system, for partitions having pressure relief openings formed therein, in nuclear reactor buildings where main coolant nozzles of a reactor pressure vessel penetrate a biological shield, comprising blowout inserts in the form of closure elements having a side facing the reactor, a side facing away from the reactor and anchors for holding said closure elements, said blowout inserts being pushable out of said anchors by an overpressure in a given pressure difference direction on said side facing the reactor, and an outer sealing blowout skin, said blowout inserts being in the form of heat-retarding cassette inserts having a front surface at said side facing away from the reactor with a peripheral shearing edge formed thereon resting against said blowout skin, said blowout skin having a given thickness in said given pressure difference direction, said blowout skin being sheared off by said shearing edge of said cassette insert and said cassette insert being pushed out of said anchors when a given pressure difference is reached, said anchors being in the form of a partition seat having an outside and an inside and said partition seat and blowout insert being widened from the inside to the outside. 2. Closure system according to claim 1, wherein said cassette insert contains retarding material in the form of mineral fibers. 3. Closure system according to claim 1, wherein said cassette insert is formed of metal and includes layered and mutually spaced apart retarding foils in the interior thereof for retarding pressure in said pressure difference direction, said foils forming retarding cells. 4. Closure system according to claim 1, wherein said blowout insert and partition seat are conically shaped from the inside to the outside. 5. Closure system according to claim 1, wherein said blowout insert and partition seat are truncated pyramoidally shaped from the inside to the outside. 6. Closure system according to claim 1, wherein said cassette insert has a partition area and said blowout skin is in the form of a glass blowout pane being clamped and sealed to said partition area. 7. Closure system according to claim 1, wherein said blowout skin is a metallic blowout foil. 8. Closure system according to claim 1, wherein said cassette insert includes a peripheral sheet metal envelope having a front surface and said shearing edge projects from said front surface of said envelope. 9. Closure system according to claim 3, wherein one of said foils is a front foil forming said shearing edge, said front foil having a doubly chamfered rim being substantially V-shaped and having a relatively greater thickness than the others of said foils. |
summary | ||
summary | ||
048809888 | description | Detailed Description The radiation arrangement is used in a test apparatus to test permanence of a test sample T (FIG. 2) with respect to influences of light and weather. A tubular cylindrical filter 1 surrounds a radiation source 2. The radiation source, preferably is an elongated tubular source having a longitudinal axis 3 which is arranged parallel to the axis 4 of the cylindrical filter 1. The radiation source is spaced from the axis 4 of the filter 1. The axis 3 of the radiation source 2 and the axis 4 of the filter 1 together define a plane in which an IR radiation accepting element defining a flat surface 5 is located. The surface element 5 subdivides, in the region 15 between radiation source 2 and the inner surface of the filter 1, the interior space of the tubular filter 1. The region 16 between the light source 2 and the other or diametrically opposite wall portion of the filter 1 can be closed off by another IR absorbing element 9 (FIG. 2) or can be left open (FIG. 1) as desired. Looked at in cross section, as seen in FIG. 2, the surface 5 is placed in the region which has the largest spacing between the radiation source and the inner surface of the filter 1, and defined by the axis 4 of the filter and the axis 3 of the radiation source. The inner surface of the filter is coated with a filter layer 6 of a dialectric material which has the characteristic that the radiation from the radiator 2 is selectively divided. The filter layer 6 is transparent for UV radiation and for visible radiation, as schematically shown by arrows 7. The layer 6 is, however, reflective for IR radiation, as schematically shown by arrows 8, which IR radiation is reflected back into the interior space of the filter 1. The reflected IR radiation impinges on the absorbing surface 5. A suitable element forming the surface 5 is a flat sheet-metal structure which is blackened. Radiation accepted by the element defining the surface 5 can thus be readily carried away as heat. It is possible to construct the element defining the surface 5 as a cooled structure, if desired. In order to further suppress IR radiation randomly reflected in the inside of the tube 1, a further absorbent surface 9, similar to surface 5, can be placed in the region 16 (FIG. 2), and co-planar with the surface 5. Preferably, the radiation source 2 is so located that its axis 3 has a distance 10 from the axis 4 of the filter 1 which is about 1.5 times the diameter 11 of the radiator source 2. Preferably the radiation source is a xenon radiator. Further the center of the axis 3 of the radiation 2 is offset at the most by 0.25 of the inner diameter of the filter 1 from the axis 4 of the filter 1. By locating the radiation source 2 eccentrically with respect to the axis 4 of the cylindrical filter 1, the entire reflected IR radiation, reflected from the inner surface of the filter 1, is directed to the absorbing surface 5 and, possibly, any stray radiation to the surface 9, if used. A source 2 may be used which, rather than being cylindrical, is essentially spherical or bulbous in shape. Such a source is schematically shown by the circle 12 (FIG. 1). Since the radiation source, as such, is well known and a standard article of commerce, it need not be further described and any suitable such source may be used. Is an essentially punctiform or bulbeous radiation source if used, it is preferably located at about half the axial height of the cylinder 1. Various changes and modification may be made within the scope of the invention concept. A suitable material for the dielectric IR reflective UV and visible transmissive coating 6 is a compound comprising SiO.sub.2, ThO.sub.2, Al.sub.2 O.sub.3, Fe.sub.2 O.sub.3 ; the material for the tubular cylindrical support is SiO.sub.2. |
059011939 | abstract | The invention relates to a nuclear fuel element for pressurized-water reactors comprising a fuel cladding of a zirconium alloy composed of two parts. The inner part constitutes a supporting part of a conventional zirconium and the outer part a corrosion-resistant layer of zirconium with 0.65-0.95% Sn and 0.4-0.5% Fe. The difference in Sn content between the inner and outer parts must not exceed 0.7%. The cladding is finally heat-treated within the range 450 to 510.degree. C. |
056174560 | abstract | The lower end of each of a plurality of fuel rods is supported by a fuel supporting portion of a lower tie plate. The fuel supporting portion includes a plurality of second coolant paths for supplying a coolant from below the fuel supporting portion to a first coolant path defined above the fuel supporting portion and between the fuel rods. The total cross-sectional area of all the second coolant paths is smaller than the cross-sectional area of the first coolant path. A water rod disposed between the fuel rods each includes an ascending tube path having therein a coolant ascending path having a coolant inlet port opening to a region below the fuel supporting portion, for guiding upward the coolant supplied through the coolant inlet port, and a descending tube having therein a coolant descending path having a coolant delivery port opening to said first coolant path, for guiding downward the coolant guided by said coolant ascending path and discharging the coolant through the coolant delivery port to the first coolant path. The descending tube path is so disposed outside the ascending tube path as to define a gap, through which the coolant inside the first coolant path flows, between it and the ascending tube path. |
description | As shown in FIG. 1, an optical element 2 includes a substrate 4 with a multilayer structure 6 applied thereto. The multilayer structure 6 includes a plurality of layer sets 8 where each layer set 8 is made up of two separate layers of different materials: one with relatively high atomic number, or Z, and a second with relatively low atomic number. As shown in FIG. 1, the layer sets 8 are uniform, meaning that the d spacing does not vary either laterally or through the depth of the multilayer structure. The thickness d of each layer set 8 has a value ranging from approximately 1 nm to approximately 100 nm depending on the energies of the radiation to be reflected. From about 10 to 1000 layer sets may be deposited on a substrate, depending on the desire qualities of the multilayer structure 6. The layer sets 8 preferably are composed of two material layers 10, 12 with diverse electron densities. Each of the material layers 10 and 12 preferably have a substantially identical thickness ranging from approximately 0.5 nm to approximately 50 nm. The absorber or high electron density layer 10 behaves like the plane of atoms in a crystal, while the low electron density layer 12 is analogous to the space between the planes. The high electron density layers 10 can include either W, Ni, Mo, Fe, Cr, Co, V, Mn, Nb, Ru, Rh, Pd, La, Ta, Re or Pt. The low electron density layer can include either silicon, carbon, B4C, Be, Li, B, Al or Sc. Thus, examples of layer sets 8 are WB4C, NiB4C, NiC, WSi, MoSi or MoB4C. Note that while the layer sets 8 of FIG. 1 show the low Z/low electron density layers 10 above the high Z/high electron density layers 12, it is possible to reverse their order without departing from the spirit of the invention so that the high Z layers 12 are above the low Z layers 10. If such reversal is performed, then the thicknesses of the various layers 10, 12 and the protective layer 14 are kept the same in the case where they are not reversed. The substrate 4 upon which the layer sets 8 are produced must meet precise specifications. The surface of substrate 4 must be capable of being polished to roughness that is precise on an atomic level. The root mean squared surface roughness of the substrate of the preferred embodiment will range from 0.5 to 20 angstroms, measured at intervals of about 10 angstroms. Examples of material used for substrate 4 are silicon wafers, mica, quartz, zerodur, sapphire, germanium, pyrex, silicon carbide or other like substances as are described in U.S. Pat. No. 5,646,976, the entire contents of which are incorporated herein by reference. In order to reduce the risk of radiation enhanced damage of the multilayer structure 6, a protective coating 14 is applied on the exterior surface 16 of the multilayer structure by such well known techniques as sputter down techniques, magnetron sputtering, e-beam evaporation, ion-beam sputtering, evaporation, electron beam implantation/plating. The protective coating 14 is generally made of a stable element like Silicon or a stable compound like SiC. The protective coating 14 has a thickness having a value ranging from approximately 60 xc3x85 to approximately 500 xc3x85, which corresponds to a loss in reflectivity of about 0.5% to 1% for Cu Kxcex1 radiation depending on the material used and the thickness of the protective coating 14. In addition, the thickness of the protective coating 14 can vary depending on the wavelength of the incident radiation. In the case of the protective coating 14 being silicon and the layer sets 8 being NiB4C, air interacts with the silicon to form a very thin layer of the oxide SiO2, which is known to be very resistant to radiation enhanced damage. The oxide reduces the possibility of the absorber layer 10 of Ni reacting with ionizing oxygen thereby extending its lifetime and the lifetime of the multilayer structure 6. The embodiment of the optical element described above with respect to FIG. 1 uses a single layer for the protective coating. As will be explained below with respect to FIGS. 2 and 3, it is possible to embody the protective coating as a multi-layer structure. In particular, an optical element 2xe2x80x2 shown in FIG. 2 includes a multilayer protective coating 14xe2x80x2 deposited on the multilayer structure 6 and substrate 4 discussed previously with respect to FIG. 1. The multilayer protective coating 14xe2x80x2 includes a plurality of layer sets 18 where each layer set 18 is made up of two separate layers of different materials: one with relatively high atomic number, or Z, and a second with relatively low atomic number. The multilayer protective coating 14xe2x80x2 has two roles to perform: 1) act as a protective coating and 2) contribute to reflection. The second role implies that the multilayer protective coating 14xe2x80x2 will have a spacing dxe2x80x2 that is the same as the spacing d of the multilayer structure 6. In particular, the layer sets 18 are uniform, wherein each layer set 18 has a spacing dxe2x80x2 that has a value ranging from approximately 1 nm to approximately 100 n depending on the energies of the radiation to be reflected. The layer sets 18 preferably are composed of two material layers 20, 22 with diverse electron densities. Each of the material layers 20 and 22 preferably has a substantially identical thickness having a value ranging from approximately 0.5 nm to approximately 50 nm. The absorber or high electron density layer 20 can include either W, Ni, Mo, Fe, Cr, Co, V, Mn, Nb, Ru, Rh, Pd, La, Ta, Re or Pt. While the high electron density layer 20 preferably is made of the same material as the high electron density layer 12 of the multilayer structure 6, it can be made of a different material. The low electron density layer 22 can include either silicon, carbon, B4C, Be, Li, B, Al or Sc, wherein the low electron density layer 22 is made of a material that is different than the material of the low electron density layer 10 of the multilayer structure 6. The low electron density layer 22 should be more resistant to radiation damage than the low electron density layer 10 and should have a density that is as close as possible to the density of the low electron density layer 10. Note that while the layer sets 8 and 18 of FIG. 2 show the low Z/low electron density layers 10, 22 above the high Z/high electron density layers 12, 20, respectively, it is possible to reverse their order without departing from the spirit of the invention so that the high Z layers 12, 20 are above the low Z layers 10, 22. If such reversal is performed, then the thicknesses of the various layers 10, 20, 12 and 22 are kept the same in the case where they are not reversed. Normally, the number of layer sets 18 of the protective coating 14xe2x80x2 is much smaller than the number of layer sets 8 of the multilayer structure 6. For example, if the number of layer sets 8 deposited for the multilayer structure 6 is, say, 80 then, the number of layer sets 18 deposited for the multilayer protective coating 14xe2x80x2 is around 10. There are several advantages for using a multilayer protective coating 14xe2x80x2 instead of a single layer protective coating 14. For example, a single layer protective coating acts primarily as an absorber. That is, while it shields the multilayer structure from direct exposure to atmosphere, it also absorbs x-rays thereby reducing the reflectivity. On the other hand, deposition of a multilayer protective coating, while shielding the multilayer structure from direct exposure to atmosphere, also contributes to reflection. This implies that the loss of reflectivity is minimized. Another advantage of using a multilayer protective coating is that since such a coating contributes to reflection, many layers can be deposited which, in turn, increases the thickness of the protective coating thereby affording better protection to the multilayer structure. A simple example of the above scheme is to deposit 80 layer sets 8 of Ni/C, which will act as the multilayer structure 6 and then deposit 10 layer sets 18 of Ni/Si on top of the multilayer structure 6, which will act as the multilayer protective coating 14xe2x80x2. A second embodiment of an optical element that includes a multilayer protective is shown in FIG. 3. The optical element 6xe2x80x3 includes a multilayer protective coating 14xe2x80x3 deposited on the multilayer structure 6 and substrate 4 discussed previously with respect to FIG. 1. The multilayer protective coating 14xe2x80x3 includes a plurality of layer sets 18xe2x80x2 where each layer set 18xe2x80x2 is made up of two separate layers of different materials, each with a relatively low atomic number. In contrast with the multilayer protective coating 14xe2x80x2 of FIG. 2, the multilayer protective coating 14xe2x80x3 does not significantly contribute to reflection. Instead, the multilayer protective coating 14xe2x80x3 is designed to increase resistance to radiation damage by exploiting resistive properties of the layer sets 18xe2x80x2. Since the multilayer protective coating 14xe2x80x3 does not significantly contribute to reflection, the spacing dxe2x80x2 of the layer sets 18xe2x80x2 may or may not be equal and need not be the same as the spacing d of the multilayer structure 6. In this scheme, a single layer topcoat is replaced with multiple layers of light elements. Consider, for example, a multilayer of NiC. A single Si layer of thickness, say 100 xc3x85, can be deposited in the embodiment of FIG. 1 or two layer sets 18xe2x80x2 can be deposited, wherein each layer set 18xe2x80x2 contains alternating 25 xc3x85 thick layers of Si and C, as shown in FIG. 3. In this scenario, the top layers do not contribute to reflection but afford the possibility of making the protective topcoat 14xe2x80x2 more resistant to radiation damage by exploiting resistive properties of their combination. In this scenario, the thickness dxe2x80x3 of the layer sets 18xe2x80x2 mayor may not be equal and need not be equal to the d spacing of the multilayer 6. The thicknesses of the layer sets 18xe2x80x2 are primarily determined by a combination that affords the best protection. The layer sets 18xe2x80x2 preferably are composed of two material layers 20xe2x80x2, 22xe2x80x2 with similar electron densities. The low electron density layers 20xe2x80x2, 22xe2x80x2 each include either silicon, carbon, B4C, Be, Li, B, Al or Sc, wherein the material of layer 20xe2x80x2 differs from the material of layer 22xe2x80x2. The thicknesses and numbers of the layer sets 18xe2x80x2 and layers 20xe2x80x2 and 22xe2x80x2 are chosen so as to maximize radiation protection for the multilayer structure 6. For example, two layer sets 18xe2x80x2 having a spacing dxe2x80x3 of 50 xc3x85 can be used, wherein each layer set 18xe2x80x2 includes alternating layers 20xe2x80x2 and 22xe2x80x2 of Si and C. The layers 20xe2x80x2 and 22xe2x80x2 have equal thicknesses of 25 xc3x85. The multilayer structures 6 and protective coatings 14, 14xe2x80x2, 14xe2x80x3 described above with respect to FIGS. 1-3 can be applied in another of ways. For example, the multilayer structure 6 and protective coatings 14, 14xe2x80x2, 14xe2x80x3 can be used to form either flat or curved optical elements 2, 2xe2x80x2, 2xe2x80x3. An example of a method of forming flat and curved optical elements is described in U.S. Pat. No. 5,646,976, the entire contents of which are incorporated herein by reference. Proposed applications of such optical elements 2, 2xe2x80x2, 2xe2x80x3 include spectroscopy and diffractometry. As shown in FIG. 4, such systems include an x-ray source 16 that emits a set of x-rays that are directed to the optical elements 2, 2xe2x80x2 or 2xe2x80x3. X-rays pass through the protective layer 14, 14xe2x80x2, 14xe2x80x3 and are diffracted by the multilayer structure 6. The optical elements 2, 2xe2x80x2, 2xe2x80x3 could also be applied to focusing optics, for x-ray lithography and microscopy, in particular, optics for high resolution scanning x-ray microscopy, point-to-point imaging optics including multi-element systems, an optic for monochromatization of broad-band radiation, synchrotron radiation in particular. Many medical applications are also contemplated, in particular, as power filters to eliminate undesired energy or use in radiography where a high contrast image is desired. While the above description constitutes the preferred embodiments of the present invention, it will be appreciated that the invention is susceptible of modification, variation and change without departing from the proper scope and fair meaning of the accompanying claims. For example, the optical elements can also be used for transformation beams of cold and thermal neutrons. In particular, they can be used for increasing density and uniformity of neutron flux or separation of the neutrons with different spin. |
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052981967 | abstract | Apparatus for heating an object such as a nuclear target bundle to release and recover hydrogen and contain the disposable residue for disposal. The apparatus comprises an inverted furnace, a sleeve/crucible assembly for holding and enclosing the bundle, conveying equipment for placing the sleeve onto the crucible and loading the bundle into the sleeve/crucible, a lift for raising the enclosed bundle into the furnace, and hydrogen recovery equipment including a trap and strippers, all housed in a containment having negative internal pressure. The crucible/sleeve assembly has an internal volume that is sufficient to enclose and hold the bundle before heating; the crucible's internal volume is sufficient by itself to hold and enclose the bundle's volume after heating. The crucible can then be covered and disposed of; the sleeve, on the other hand, can be reused. |
claims | 1. A wearable radiation detection system comprising:a housing;one or more pixelated imager chips located within the housing, each of the one or more pixelated imager chips being configured to operate as a direct detector, wherein the one or more pixelated imager chips comprise one or more of a photodiode, a color imager, a monochrome imager, a low light imager, an infrared imager, a thermal imager, a complementary metal-oxide-semiconductor imager, and a charge-coupled device imager; andone or more processors operably connected to each of the one or more pixelated imager chips, the one or more processors being configured to detect ionizing radiation using image data. 2. The wearable radiation detection system of claim 1, wherein the one or more pixelated imager chips are configured to acquire the image data without a scintillator. 3. The wearable radiation detection system of claim 2, wherein the image data comprises two-dimensional image data. 4. The wearable radiation detection system of claim 3, further comprising:one or more position sensors operably connected to the one or more processors and configured to acquire position data. 5. The wearable radiation detection system of claim 4, wherein the one or more processors are further configured to calculate a three-dimensional mass distribution based on the two-dimensional image data and the position data. 6. The wearable radiation detection system of claim 1, further comprising:one or more position sensors operably connected to the one or more processors and configured to acquire position data. 7. The wearable radiation detection system of claim 1, wherein the housing comprises one or more of a badge, a patch, a bracelet, a pendant, or a wrist computing device. 8. The wearable radiation detection system of claim 1, wherein the one or more processors are further configured to record a time stamp. 9. The wearable radiation detection system of claim 1, further comprising:a transmitter in operable communication with the one or more processors, wherein the one or more processors are further configured to cause the transmitter to wirelessly transmit the image data to a remote computing device. 10. The wearable radiation detection system of claim 1, further comprising:one or more body temperature sensors configured to detect a temperature of a body of a user. 11. A wearable radiation detection system comprising:a wearable housing;one or more pixelated imager chips configured to operate as direct detectors and located within the wearable housing;a processor operably connected to each of the one or more pixelated imager chips; and a non-transitory, processor readable storage medium operably connected to the processor, wherein the storage medium contains one or more instructions that, when executed, cause the processor to: receive image data from the one or more pixelated imager chips, and detect ionizing radiation based on the image data. 12. The wearable radiation detection system of claim 11, wherein:the one or more pixelated imager chips comprise one or more of a photodiode, a color imager, a monochrome imager, a low light imager, an infrared imager, a thermal imager, a complementary metal-oxide-semiconductor imager, and a charge-coupled device imager; andthe one or more pixelated imager chips are configured to acquire the image data without a scintillator. 13. The wearable detection system of claim 11, wherein one or more pixelated imager chips comprise two or more pixelated imager chips positioned in two or more different orientations. 14. The wearable detection system of claim 13, wherein the two or more pixelated imager chips comprise a first pixelated imager chip of a first type and a second pixelated imager chip of a second type different from the first type. 15. The wearable detection system of claim 13, wherein the wearable housing comprises one or more of a badge, a patch, a bracelet, a pendant, or a wrist computing device. 16. A wearable radiation detection system comprising:two or more pixelated imager chips configured to operate as direct detectors;one or more processors;memory in communication with the one or more processors and storing instructions that, when executed, are configured to cause the system to:calibrate the two or more pixelated imager chips to detect radiation;capture, using the two or more pixelated imager chips, image data in a manner controlled by the processor;process the image data; anddetect radiation based on the processed image data. 17. The wearable radiation detection system of claim 16, wherein:the two or more pixelated imager chips are positioned in two or more different orientations; andthe manner controlled by the processor is a long exposure image or video. 18. The wearable radiation detection system of claim 16, wherein the two or more pixelated imager chips are of two or more different types, and each of the two or more types of the two or more pixelated imager chips is calibrated differently. 19. The wearable radiation detection system of claim 16, wherein the radiation is detected in real-time, and the instructions are further configured to cause the system to trigger one or more of an alert indicative of detected radiation deviating from a threshold level or a notification indicative of a measured radiation dose rate. 20. The wearable radiation detection system of claim 16, wherein the two or more pixelated imager chips are positioned for wearing in two or more different locations on a body of a person. |
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063226103 | summary | FIELD OF THE INVENTION This invention concerns an integrated device to inject oxygen and technological gases in general, and also solid material in powder form, as set forth in the relative main claim. The invention also concerns the method to use the integrated device for the metallurgical processing of a bath of molten metal. In the following description we shall refer principally to oxygen, but it is implicit that the device according to the invention can be used also for the injection of any other technological gas, for example air, argon, nitrogen or others, or a mixture thereof, in a metallurgical process which includes a bath of molten metal. The device according to the invention is suitable to inject, together with the gassy jet of oxygen or other technological gases, a jet of solid material in powder form, such as for example lime powder, carbon powder or similar. The invention is applied, preferably but not exclusively, in steel working industries and is particularly indicated to be applied in processes to melt steel, for example in electric arc furnaces. BACKGROUND OF THE INVENTION It is common practice in electric arc furnaces, and in other applications in steel and metallurgical industries, to inject technological gases and solid material in powder form above and inside the bath of melting metal. Generally speaking, technological gases should be taken to mean oxygen, nitrogen, argon, methane, propane, air or other gases with like characteristics. Generally speaking, solid material in powder form should be taken to mean the powders of coke, carbon, iron alloys, lime, dolomite or other materials with like characteristics. There are many purposes for this injection, among the most important being: to provide an energy input for melting; PA1 to activate the oxidation step and influence the dephosphorization and the desulphurization of the liquid metal; PA1 to encourage slag foaming; PA1 to facilitate the melting and shearing of the scrap; PA1 to encourage the stirring of the bath, thus accelerating the time taken to activate the chemical reactions; PA1 to actuate decarburation, that is to say, the regulation of the carbon content, and to control the tapping temperature; PA1 to obtain the burner functioning mode, wherein the oxygen or air enriched with oxygen act as comburents of natural gas, methane, oil, propane, butane, carbon or other solid or gassy fuels. PA1 a) by means of water-cooled lances equipped at the ends with a nozzle, for example of the convergent/divergent type, suitable to produce a supersonic jet at outlet; PA1 b) by using lances of the consumable type. PA1 the need for manipulation; PA1 the gassy jet loses energy due to the distance from the bath, which often implies it is impossible for the jet to penetrate inside the bath; PA1 large quantities of cooling water are needed to prevent the tip of the lance from being destroyed due to the heat and the mechanical stresses; PA1 danger of explosions caused by possible water leakages; PA1 if the lance is very close to the bath there is a risk of damage due to the heat, the tip may be washed, encrustations of steel may form and there is a risk of erosion; PA1 if the lance is very close to the scrap, the gassy jet may be deflected and even reflected against the end or the side of the lance and therefore cause damage thereto; PA1 another lance is needed for the combustion of the CO which escapes from the bath; PA1 other devices are needed to inject the solid material in powder form. PA1 high cost of the parts which are consumed; PA1 it is difficult to determine the exact positioning of the tip of the lance; PA1 the method is not very efficient at distributing the point of impact of the jet; PA1 due to overheating the lance may bend; PA1 it is necessary to add new segments of lance as it is gradually consumed; this requires a wide use of equipment and manipulators which are costly and bulky; PA1 further devices are needed for the post-combustion and injection of the solid material in powder form. PA1 reduction of tap-to-tap time; PA1 reduction of electric energy consumed; PA1 reduction of electrode consumption; PA1 improved penetration of the gassy jet into the bath of metal; PA1 more accentuated turbulence in the bath, which entails a more uniform temperature and a quicker melting of the scrap; PA1 greater melting intensity; PA1 greater productivity and greater efficiency in the use of the oxygen in the bath; PA1 reduced concentration of oxygen in the liquid bath and therefore better quality of steel; PA1 when coupled with electromagnetic stirrers, in some cases it allows to eliminate the function of the bottom tuyeres; PA1 reduction of erosion of the refractory; PA1 greater efficiency in post-combustion and reduction of the carbon oxide in the gases discharged from the furnace; PA1 reduction of the water cooling of the injection means; PA1 more efficient use of foamy slag technique. In the state of the art two main solutions are adopted to inject oxygen or other gases inside a liquid bath: Using water-cooled lances entails the following disadvantages: The disadvantages of consumable lances are as follows: Of the two solutions, in recent years the use of supersonic lances has particularly developed; with these it is possible to inject the necessary quantity of oxygen by means of a jet with a speed higher than that of the sonic speed of the fluid in the relevant conditions of supply temperature and pressure. However, with present-day technology, the oxygen is not injected an optimum manner of functional to the melting process. In fact, in systems known to the stat of the art, the impulse of the jet of oxygen is insufficient to penetrate the bath of liquid metal to a depth sufficient to ensure that the oxygen is adequately distributed throughout the bath (for example equal to half the overall height of the bath). At the moment of impact with the surface of the bath, the jet generates impact waves of compression of very high intensity, which cause a dissipation of the jet and a dispersion of the gas on the surface of the bath, so that only a minimum part of the gas penetrates into the liquid bath of molten metal. This impact against the surface of the bath also causes a loss of coherence and parallelism in the fluid threads in the jet, with a resultant loss in its penetrative ability. Moreover, the system makes it necessary to mount the supersonic lance on a manipulator, or another mechanical organ which allows the lance to be moved, in order to adjust the outlet distance with respect to the surface of the bath, since the jet of oxygen tends to disperse after a few centimeters, in the order of a few dozen cm, from the outlet of the supersonic nozzle. For this reason, consolidated practice provides to insert the end part of the supersonic lance inside the layer of slag above the bath to ensure that the oxygen is introduced inside the liquid bath in a sufficiently efficient manner, but in any case this is not an optimum solution. EP-A-874.194, which discloses the pre-characterizing part of claim 1, describes a burner which can be used on electric arc furnaces comprising a first, inner nozzle with a convergent-divergent development (Laval nozzle) which emits a mixture of oxygen and natural gas, and a second nozzle, coaxial to and outside the first, which emits particulate material. In this document, the purpose is substantially to allow the flow of particulate material to mix with the primary flow of oxygen and fuel, so that the material can be distributed uniformly in the flame produced by the burner and can be projected as far as possible inside the furnace. The outer nozzle defines a straight flow path for the particulate material in order to prevent abrasions on the wall due to the passage of said material. In this document, the flow delivered by the outer nozzle does not form a protective crown for the primary flow delivered by the inner nozzle, but mixes immediately therewith, already inside the burner itself, since it is drawn by the high increase in pressure created by the supersonic acceleration of the oxygen and fuel. Moreover, this document does not provide variable working options to modify the composition and the development of the flame according to the various steps of the melting process, so that the regulation of the working of the burner is not correlated to the development of the melting cycle and to the different technological requirements which gradually occur. The present Applicant has designed, tested and embodied this invention to overcome all these shortcomings and to obtain further advantages. SUMMARY OF THE INVENTION This invention is set forth and characterized in the respective main claims, while the dependent claims describe other characteristics of the main embodiment. The purpose of the invention is to achieve an integrated device to inject oxygen and technological gases into a bath of liquid metal which will ensure maximum efficiency and yield, minimum wear and minimum difficulty in use and manipulation. Another purpose of the invention is to integrate in a single device the following functions: to inject gases, to inject solid fuel in powder form or in particles, to add comburent in the post-combustion process, to inject powders to passivate the slag, and also to function as a burner; these various functions may be carried out in succession by the same device according to the progress of the melting cycle. With the device according to the invention, in fact, a single module allows to achieve all the steps of the melting process, that is to say, heating the material to be melted, melting and shearing the scrap by injecting high density oxygen, decarburation, slag foaming, injecting solid fuel in powder form or in particles, post-combustion and energy input in the decarburation step. All or part of the above-mentioned functions can be performed at the same time. In other words, according to the needs and the specific step of the cycle, the device according to the invention can be made to function either as a burner with a variable stechiometric ratio and a variable flame length, or also as an injector of oxygen only or other technological gas with a high rate of distribution of the gassy flow into the liquid bath, or also as a simultaneous injector of technological gases and powdered fuel, for example to obtain the production of foamy slag or to obtain the passivation of the slag itself. The device according to the invention therefore allows, in a single body, to integrate the function which in the state of the art are normally performed by three different devices: a supersonic lance to inject a gassy jet, a subsonic lance or burner for post-combustion, and a lance to inject solid fuel such as carbon powder or combustible powders in general. Using the device according to the invention allows to obtain substantially operating advantages, such as for example the reduction of the work force required, greater safety for the workers, improved working conditions and a greater control over the process. In terms of the process, using the device according to the invention allows to obtain the following improvements: The device according to the invention allows the simultaneous emission of two jets substantially autonomous and independent, which ensures a good working flexibility and versatility. The device according to the invention is mounted in an axially fixed manner on the wall of the furnace, and therefore does not need any manipulation, replacement of tubes, insertion of equipment through apertures in the furnace, with the consequent need of keeping the slag door open. The device is mounted in such a way that its angle of inclination with respect to the surface of the liquid bath can be varied so as to adapt it to the changing conditions which occur inside the furnace during the cycle. For example, the inclination of the device can be varied during the process in order to reduce the level of the liquid bath so as to maintain substantially constant the angle of incidence of the jet, thus preventing the unwanted phenomenon of the jet itself being reflected. The device according to the invention can be used in a modular system which provides a plurality of injection points distributed on the periphery of the furnace, which makes the injection system extremely flexible and versatile and reduces the risks of splashes of liquid metal against the walls of the furnace thanks to the fact that it is possible to use lesser smaller amounts of gas for each device. The device according to the invention allows to emit an extremely energetic gassy jet with a high specific density, which encourages the penetration into the liquid metal and increases efficiency also when there is a smaller quantity of gas introduced. |
abstract | There is provided an exit window for an electron beam from a linear accelerator for use in producing radioisotopes. The exit window comprises a cylindrical channel operatively connectable at one end to a vacuum chamber configured for travel of the electron beam; and a domed dished head at the other end of the channel, the dished head comprising a convex portion having a protruding crown configured for pass-through of the electron beam wherein the geometry of the domed dished head is proportioned to resist pressure stress created by cooling medium circulating around the protruding crown and the vacuum in the cylindrical channel and to maintain the combined cooling medium pressure stress and pulsed electron beam thermal stress below the fatigue limit of the material forming the exit window. |
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048246344 | abstract | A fuel element for a nuclear reactor having a zirconium-tin alloy cladding tube, with a thin coating of particles of enriched boron-containing compound burnable poison particles, such as ZrB.sub.2 deposited from a liquid suspension which includes an acrylic polymer binder material and isopropanol. Graphite lubricant particles can be included in the coating. |
052689433 | claims | 1. A nuclear reactor having a reactor vessel disposed in a containment shell, a normally pressurized coolant circuit including the reactant vessel, a refueling water storage tank, and a coolant addition system operable to depressurize the coolant circuit for adding coolant from the refueling water storage tank to the coolant circuit at reduced pressure, the reactor comprising: a residual heat removal loop having at least one pump and at least one heat exchanger, the residual heat removal loop having an inlet and an outlet both coupleable to the coolant circuit, the pump being operable in a normal mode to extract coolant from the coolant circuit and to move the coolant through the heat exchanger, the coolant being returned to the coolant circuit through the outlet of the residual heat removal loop; a controllable valving arrangement coupled between the inlet of the residual heat removal loop and the refueling water storage tank, the controllable valving arrangement including at least one coupling valve operable in conjunction with the residual heat removal loop for adding water from the refueling water storage tank to the coolant circuit; and an automatic depressurization system having a plurality of conduits including depressurization valves, coupling the coolant circuit and the containment shell, the depressurization valves being operable to open the conduits at successively lower levels of the high pressure makeup tank in stages of depressurization, and further comprising control means operable to activate the residual heat removal loop at one of the stages. a reactor vessel disposed in a containment building, for heating a coolant at an operational pressure via nuclear fuel, the reactor vessel being coupled via a hot leg conduit and a cold leg conduit for generating electrical power when the reactor is operational, said reactor vessel and conduits being included in a primary coolant circuit of the reactor; a high pressure makeup water storage tank selectively coupleable to the primary coolant circuit via makeup supply valves, the high pressure makeup water storage tank holding a supply of makeup coolant substantially at the operational pressure; a refueling water storage tank at ambient pressure of the containment building, the refueling water storage tank being selectively coupleable to the coolant circuit for adding coolant upon depressurization of the primary coolant circuit; a depressurization system operable to vent the primary coolant circuit into the containment building in a plurality of depressurization stages characterized by progressively reduced amounts of coolant in the supply of makeup coolant; a residual heat removal system including at least one heat exchanger and at least one pump, the residual heat removal system being normally operable following depressurization of the primary coolant circuit for removing residual heat from the primary coolant circuit when the reactor is not operational; and, wherein the residual heat removal system is selectively coupleable to pump coolant from the refueling water storage tank to the primary coolant circuit during at least one of the depressurization stages, whereby the residual heat removal system is operable selectively to increase a quantity of coolant in the primary coolant circuit at a pressure above the ambient pressure in the containment building. 2. The reactor according to claim 1, wherein the refueling water storage tank is maintained at atmospheric pressure and the controllable valving arrangement and the residual heat removal system are operable to add said water to the coolant circuit at a pressure above said atmospheric pressure. 3. The reactor according to claim 2, wherein the residual heat removal system includes at least one pump, and wherein an output of the pump is coupleable by the valving arrangement to a water injection port of the reactor vessel. 4. The reactor according to claim 1, wherein said stages of depressurization are triggered as a function of a level of coolant in the coolant circuit, and wherein one of the conduits openable during a final stage of depressurization couples the coolant circuit directly to the containment shell, whereby activation of the residual heat removal loop forestalls said final stage by addition of coolant. 5. The reactor according to claim 4, further comprising a pressurizer tank coupled to the coolant circuit, and wherein at least one initial stage of depressurization is accomplished by opening an initial stage depressurization valve coupling the pressurizer tank to a sparger in the refueling water storage tank. 6. The reactor according to claim 5, further comprising a high pressure makeup water storage tank having a makeup water supply at substantially a pressure of the coolant circuit, and further comprising sensing means associated with the high pressure makeup water storage tank for sensing a level of coolant available in the high pressure makeup water storage tank, and wherein the depressurization valves are responsive to the sensing means. 7. A pressurized water nuclear reactor, comprising: 8. The nuclear reactor according to claim 7, wherein the depressurization system comprises a pressurizer tank coupled to the primary coolant circuit and a plurality of conduits coupled to spargers in the refueling water storage tank, the conduits having depressurization valves for opening the conduits during the stages of depressurization, thereby successively increasing coupling between the primary coolant circuit and the containment building, the residual heat removal system being operable to preclude depressurization to a final one of the stages of depressurization by addition of coolant from the refueling water storage tank. 9. The nuclear reactor according to claim 8, wherein in a final stage of depressurization the containment building is flooded from the refueling water storage tank, said residual heat removal system being operable to preclude flooding of the containment building during depressurization by enabling addition of coolant above the ambient pressure in the containment building. 10. The nuclear reactor according to claim 7, wherein the residual heat removal system includes at least two parallel legs, each of the legs including a pump and a heat exchanger. 11. The nuclear reactor according to claim 7, further comprising a heat exchanger in the refueling water storage tank controllably coupleable to a heat source of the reactor, and wherein the residual heat removal system is operable to extract heat from the refueling water storage tank as a means for removing heat from said source. 12. The nuclear reactor according to claim 11, wherein the heat source is the reactor vessel. |
claims | 1. A system comprising:an article of personal protection equipment (PPE) that can be worn by a user, and at least one component coupled to the article of PPE;a smart tag coupled to the at least one component or the article of PPE;a computing device comprising one or more computer processors and a memory that further comprises instructions that when executed by the one or more computer processors cause the one or more computer processors to:receive sensed data about a working environment using one or more sensors, the sensed data being sensed apart from the component and prior to actually using the component in the working environment;determine a duration of usage of the component based on tracked data received from the smart tag; anddetermine a condition of the component by comparing the duration of usage of the component to at least one predetermined criterion, wherein the predetermined criterion is based at least in part on the sensed data about hazardous materials in the working environment and on a service life of the component. 2. The system of claim 1, wherein the instructions that cause the one or more computer processors to determine the condition of the component further comprise instructions that when executed by the one or more computer processors cause the one or more computer processors to determine the condition of the component based at least in part on additional data corresponding to the user from the smart tag that indicates at least one of identity, medical, fit test, job description, seniority, training, or qualification information of the user. 3. The system of claim 1, wherein the tracked data indicates at least one of: an entry time when the user entered the working environment, an exit time when the user exited the working environment, an entry location where the user entered the working environment, an exit location where the user exited the working environment, a type of personal protection equipment or type of component, historical information relating to the article of PPE or component, an identifier of the user, a location where the article of PPE was used, a condition under which the article of PPE was used, maintenance performed on the article of PPE, a requirement for using the article of PPE, or a description of the working environment. 4. The system of claim 1 further comprising at least one smart tag scanner configured to read the smart tag, wherein to determine a duration of usage of the component, the at least one smart tag scanner is communicatively coupled to the computing device and configured to receive the tracked data from the smart tag and send the tracked data to the computing device. 5. The system of claim 1, wherein the instructions that cause the one or more computer processors to determine the condition of the component by comparing the duration of usage of the component to at least one predetermined criterion further comprise instructions that determine whether the duration of usage of the component exceeds the service life of the component. 6. The system of claim 1, wherein the memory further comprises instructions that when executed by the one or more computer processors cause the one or more computer processors to output a notification that the at least one predetermined criterion has been satisfied. 7. The system of claim 1, wherein the memory further comprises instructions that when executed by the one or more computer processors cause the one or more computer processors to perform, based on the determination of the condition of the component, one or more operations. 8. A method comprising:providing at least one component coupled to an article of PPE that can be worn by a user;providing a smart tag coupled to at least the component or the article of PPE;receiving sensed data about a working environment using one or more sensors, the sensed data being sensed apart from the component and prior to actually using the component in the working environment;determining a duration of usage of the component based on tracked data received from the smart tag; anddetermining, by a computing device, a condition of the component by comparing the duration of usage of the component to at least one predetermined criterion, wherein the predetermined criterion is based at least in part on the sensed data about hazardous materials in the working environment and on a service life of the component. 9. The method of claim 8, wherein determining the condition of the component further comprises determining the condition of the component based at least in part on additional data corresponding to the user from the smart tag that indicates at least one of identity, medical, fit test, job description, seniority, training, or qualification information of the user. 10. The method of claim 8, wherein the tracked data indicates at least one of: an entry time when the user entered the working environment, an exit time when the user exited the working environment, an entry location where the user entered the working environment, an exit location where the user exited the working environment, a type of personal protection equipment or type of component, historical information relating to the article of PPE or component, an identifier of the user, a location where the article of PPE was used, a condition under which the article of PPE was used, maintenance performed on the article of PPE, a requirement for using the article of PPE, or a description of the working environment. 11. The method of claim 8, wherein at least one smart tag scanner is configured to read the smart tag, wherein determining a duration of usage of the component further comprises: receiving, by the computing device and from the at least one smart tag scanner communicatively coupled to the computing device, the tracked data from the smart tag. 12. The method of claim 8, wherein determining the condition of the component by comparing the duration of usage to at least one predetermined criterion further comprises determining whether the duration of usage of the component exceeds the service life of the component. 13. The method of claim 8, further comprising outputting a notification that the at least one predetermined criterion has been satisfied. 14. A computing device comprising:one or more computer processors; anda memory comprising instructions that when executed by the one or more computer processors cause the one or more computer processors to:receive sensed data about the working environment from one or more sensors, the sensed data being sensed apart from a component coupled to an article of PPE that can be worn by a user and prior to actually using the component in the working environment, wherein a smart tag is coupled to at least the component or the article of PPE;determine a duration of usage of the component based on tracked data received from the smart tag; anddetermine a condition of the component by comparing the duration of usage of the component to at least one predetermined criterion, wherein the predetermined criterion is based at least in part on the sensed data about hazardous materials in the working environment and on a service life of the component. 15. The computing device of claim 14, wherein the memory comprises instructions that when executed by the one or more computer processors cause the one or more computer processors to determine the condition of the component based at least in part on additional data corresponding to the user from the smart tag that indicates at least one of identity, medical, fit test, job description, seniority, training, or qualification information of the user. 16. The computing device of claim 14, wherein the tracked data indicates at least one of: an entry time when the user entered the working environment, an exit time when the user exited the working environment, an entry location where the user entered the working environment, an exit location where the user exited the working environment, a type of personal protection equipment or type of component, historical information relating to the article of PPE or component, an identifier of the user, a location where the article of PPE was used, a condition under which the article of PPE was used, maintenance performed on the article of PPE, a requirement for using the article of PPE, or a description of the working environment. 17. The computing device of claim 14, wherein the memory comprises instructions that when executed by the one or more computer processors cause the one or more computer processors to receive, from at least one smart tag scanner communicatively coupled to the computing device, the tracked data from the smart tag. 18. The computing device of claim 14, wherein the memory comprises instructions that when executed by the one or more computer processors cause the one or more computer processors to determine whether the duration of usage of the component exceeds the service life of the component. 19. The computing device of claim 14, wherein the memory comprises instructions that when executed by the one or more computer processors cause the one or more computer processors to output a notification that the at least one predetermined criterion has been satisfied. 20. The computing device of claim 14, wherein the memory comprises instructions that when executed by the one or more computer processors cause the one or more computer processors to perform, based on the determination of the condition of the component, one or more operations. |
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abstract | The invention comprises a method and apparatus for imaging a tumor of a patient using one or more imaging systems positionable about the tumor and treating the tumor using positively charged particles, such as a process of: (1) rotating a gantry support and/or gantry, connected to at least a portion of a beam transport system configured to pass a charged particle treatment beam, circumferentially about the patient and a gantry rotation axis; (2) translating a translatable imaging system past the patient on a path parallel to an axis perpendicular to the gantry rotation axis; (3) imaging the tumor using the translatable imaging system; and (4) treating the tumor using the rotatable treatment beam. |
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summary | ||
claims | 1. A device for accelerating a particle beam to an energy for irradiating a target volume, the device comprising:a particle accelerator operable in a first working phase and a second working phase, particles of the particle beam being accelerated to the energy in the first working phase and the particles of the particle beam being provided and extracted for irradiating the target volume in the second working phase; anda control device configured to:monitor a motion of the target volume;interrupt an irradiation of the target volume when the target volume assumes a predetermined position or orientation that is determinable based on the monitored motion; andcontrol the particle accelerator as a function of a comparison between a residual particle number stored in the particle accelerator and a reference value. 2. The device as claimed in claim 1, wherein the control device is operable to switch the particle accelerator from the second working phase to the first working phase if the residual particle number is less than the reference value. 3. The device as claimed in claim 2, wherein the reference value used for the comparison is determined from a residual particle number that is required for irradiating a sub-region of the target volume. 4. The device as claimed in claim 1, wherein the control device is operable to discard the residual particle number stored in the particle accelerator if the residual particle number is less than the reference value. 5. The device as claimed in claim 2, wherein the control device is operable to discard the residual particle number stored in the particle accelerator if the residual particle number is less than the reference value. 6. The device as claimed in claim 4, wherein the reference value used for the comparison is determined from a residual particle number that is required for irradiating a sub-region of the target volume. 7. The device as claimed in claim 6, wherein the sub-region of the target volume is an area which is to be irradiated with the same particle energy. 8. The device as claimed in claim 1, wherein the reference value used for the comparison is determined from a residual particle number that is required for irradiating a sub-region of the target volume. 9. The device as claimed in claim 8, wherein the sub-region of the target volume is an area which is to be irradiated with the same particle energy. 10. A method for irradiating a target volume, the method comprising:accelerating particles of a particle beam to an irradiation energy provided for the irradiation of a target volume in a first working phase of a particle accelerator;irradiating the target volume using the accelerated particles in a second working phase of the particle accelerator;monitoring a motion of the target volume;interrupting the second working phase of the particle accelerator when the target volume assumes a predetermined position or orientation that is determinable based on the monitored motion;comparing, following the interrupting, a remaining residual particle number stored in the accelerator with a reference value; andcontrolling the particle accelerator as a function of a result of comparing the remaining residual particle number with the reference value. 11. The method as claimed in claim 10, wherein controlling comprises switching the particle accelerator from the second working phase to the first working phase if the residual particle number is less than the reference value. 12. The method as claimed in claim 11, wherein controlling comprises discarding the residual particle number stored in the particle accelerator if the residual particle number is less than the reference value. 13. The method as claimed in claim 10, wherein controlling comprises discarding the residual particle number stored in the particle accelerator if the residual particle number is less than the reference value. 14. The method as claimed in claim 10, wherein comparing comprises determining the reference value based on a residual particle number that is required for irradiating a sub-region of the target volume. 15. The method as claimed in claim 14, wherein the sub-region of the target volume is an area which is provided for irradiation with the same particle irradiation energy. 16. The method as claimed in claim 11, wherein comparing comprises determining the reference value based on a residual particle number that is required for irradiating a sub-region of the target volume. 17. The method as claimed in claim 16, wherein the sub-region of the target volume is an area which is provided for irradiation with the same particle irradiation energy. 18. The method as claimed in claim 13, wherein comparing comprises determining the reference value based on a residual particle number that is required for irradiating a sub-region of the target volume. 19. The method as claimed in claim 18, wherein the sub-region of the target volume is an area which is provided for irradiation with the same particle irradiation energy. 20. The device as claimed in claim 1, wherein the control device is configured to control the particle accelerator as a function of the comparison following interruption of the irradiation. |
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claims | 1. An ion implant apparatus comprising:a rotary scan assembly having an axis of rotation and a periphery;a plurality of substrate holders distributed about said periphery, said substrate holders arranged to hold respective planar substrates at a common substrate tilt angle to define a total cone angle about said axis of rotation which is less than 60°;and a beam line assembly to provide a beam of ions for implantation in said planar substrates on said substrate holders, said beam line assembly comprising in sequence in an ion beam direction: an ion source, an ion accelerator effective to accelerate ions from said ion source to produce an accelerated beam having a desired implant energy of at least 500 keV, and a beam bending magnet having a beam inlet to receive said accelerated beam; wherein said ion source, said accelerator and said beam inlet defining a beam acceleration path which is linear from said ion source to said beam inlet of said beam bending magnet;wherein said beam line assembly is arranged to direct said beam in a predetermined ion implant direction along a final beam path, which is at an angle of at least 45° to said axis of rotation; andwherein said planar substrates on said substrate holders successively intercept said final beam path in a travel direction as said rotary scan assembly rotates. 2. Ion implant apparatus as claimed in claim 1, wherein said beam line assembly includes an analyzer magnet which is located after said beam acceleration path and is operative to produce an angular separator in said accelerated beam between ions of different mass/charge ratios (m/e). 3. Ion implant apparatus as claimed in claim 2,wherein said analyzer magnet is operative to direct ions having an m/e which is desired for implantation in said predetermined ion implant direction along said final beam path,wherein said implant apparatus further comprises an ion beam dump which is mounted on said rotary scan assembly and forms an annular beam dump region which rotates with said rotary scan assembly, andwherein said analyzer magnet is operative to direct ions having an m/e greater than said desired m/e towards said annular beam dump region. 4. Ion implant apparatus as claimed in claim 1, wherein said beam bending magnet is a beam scanner magnet operative to deflect said accelerated beam at a repetition rate through a range of deflection angles, to produce a scanned beam such that said final beam path is scanned transversely relative to said travel direction of said substrate holders. 5. Ion implant apparatus as claimed in 4, wherein said beam line assembly includes an analyzer magnet which is located to receive said scanned beam from said beam scanner magnet over said range of deflection angles and is operative to produce an angular separation between beam ions of different mass/charge ratios (m/e). 6. Ion implant apparatus as claimed in 5, wherein said analyzer magnet is operative to direct ions having an m/e which is desired for implantation in a collimated scanned beam in said predetermined ion implant direction along said final beam path. 7. Ion implant apparatus as claimed in claim 6, further comprising an ion beam dump mounted on said rotary scan assembly and forming an annular beam dump region which rotates with said rotary scan assembly, wherein said analyzer magnet is operative to direct ions of said accelerated beam having an m/e greater than said desired m/e towards said annular beam dump region. 8. Ion implant apparatus as claimed in claim 1, wherein said final beam path has a total ion drift distance which is less than a diameter of said periphery of said rotary scan assembly. 9. Ion implant apparatus as claimed in claim 8, wherein said total drift distance is less than half of said diameter. 10. Ion implant apparatus as claimed in claim 1, wherein said acceleration path is aligned with said axis of rotation of said rotary scan assembly. 11. A method of implanting ions into planar substrates mounted on substrate holders distributed around a periphery of a rotary scan assembly, in which said substrate holders hold said planar substrates at a common substrate tilt angle to define a total cone angle about an axis of rotation of said scan assembly which is less than 60°, the method comprising the steps of:a) generating a source of ions including ions desired for implantation;b) extracting and accelerating said ions from said source along a linear acceleration path to produce an accelerated ion beam having an energy of at least 500 keV;c) bending said accelerated beam to direct an accelerated beam of said ions desired for implantation in a predetermined implant direction along a final beam path which is at an angle to said axis of rotation of at least 45°, andd) rotating said rotary scan assembly such that said substrates successively intercept said beam path in a travel direction. 12. A method of implanting ions as claimed in claim 11, further comprising the step of analyzing said accelerated beam using an analyzer magnet to provide an angular separation between ions of different mass/charge ratio (m/e). 13. A method of implanting ions as claimed in claim 12, wherein said analyzer magnet directs ions having an m/e which is desired for implantation in said predetermined ion implant direction along said final beam path, and wherein said analyzer magnet directs ions having an m/e greater than said desired m/e towards an annular beam dump region of a beam dump. 14. A method of implanting ions as claimed in claim 11, further comprising the step of deflecting said accelerated beam at a repetition rate through a range of deflection angles, to produce a scanned beam such that said final beam path is scanned transversely relative to said travel direction of said substrates. 15. A method of implanting ions as claimed in claim 14, further including the steps of:receiving said scanned beam over said range of deflection angles in an analyzer magnet; andanalyzing said scanned beam using said analyzer magnet to produce an angular separation between beam ions of different mass/charge ratios (m/e). 16. A method of implanting ions as claimed in claim 15, wherein said analyzer magnet directs ions having an m/e which is desired for implantation in a collimated scanned beam in said predetermined ion implant direction along said final beam path. 17. A method of implanting ions as claimed in claim 16, wherein said analyzer magnet directs ions having an m/e greater than said desired m/e towards an annular beam dump region of an ion beam dump mounted on said rotary scan assembly. 18. A method of implanting ions as claimed in claim 11, wherein ions to be implanted drift along said final beam path over a total ion drift distance which is less than a diameter of said periphery of said rotary scan assembly. 19. A method of implanting ions as claimed in claim 18, wherein said total ion drift distance is less than half of said diameter. 20. A method of implanting ions as claimed in claim 11, wherein said linear acceleration path is aligned with said axis of rotation of said rotary scan assembly. |
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abstract | Systems, devices, methods, and compositions are described for providing an x-ray shielding system including a flexible layer including a support structure having a plurality of interconnected interstitial spaces that provide a circulation network for an x-ray shielding fluid composition. |
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description | This application claims the benefit of priority from Chinese Patent Application No. 202111214851.2, filed on Oct. 19, 2021. The content of the aforementioned application, including any intervening amendments thereto, is incorporated herein by reference in its entirety. This application relates to manufacturing of vacuum chambers of nuclear fusion reactors, and more particularly to a method, device and apparatus for machining a groove of poloidal segments of a vacuum chamber of a fusion reactor, and a computer-readable storage medium. A vacuum chamber of a nuclear fusion reactor is generally divided into eight 45° sector regions (each accounting for ⅛ of the vacuum chamber) along a circumferential direction. These ⅛ sector regions are separately fabricated and then transported to the assembly site to be assembled into a 360° toroidal vacuum chamber, where each ⅛ sector region is composed of two symmetrical 1/16 sector regions; each 1/16 sector region is composed of two symmetrical 1/32 sector regions; and each 1/32 sector region is divided into four poloidal segments along a poloidal direction. However, the existing manufacturing process of poloidal segments generally struggles with large forming and assembly errors, and obvious welding deformation, such that the machining allowance of the poloidal segments cannot be accurately determined, greatly lowering the machining precision of welding grooves. An object of the present disclosure is to provide a method, device, and apparatus for machining a groove of a poloidal segment of a vacuum chamber in a nuclear fusion reactor, and a computer-readable storage medium to overcome the defects in the prior art that it fails to ensure the machining precision of welding grooves since the machining allowance of the poloidal segments cannot be accurately determined during the fabrication process. Technical solutions of the present disclosure are described as follows. In a first aspect, the present disclosure provides a method for machining a groove of poloidal segments of a vacuum chamber of a fusion reactor, comprising: collecting three-dimensional (3D) point cloud data of a surface of individual poloidal segments of the vacuum chamber of the fusion reactor; performing a reverse model reconstruction, based on the three-dimensional point cloud data, to generate an actual 3D model of the vacuum chamber of the fusion reactor; and acquiring a sectional view of the vacuum chamber of the fusion reactor according to the actual 3D model; and extracting a cross-reconstruction region between two adjacent poloidal segments from the sectional view; calculating a target machining allowance of each of the poloidal segments according to the cross-reconstruction region and a preset segment boundary; generating a machining strategy for a groove of each of the poloidal segments according to the target machining allowance and a target groove size; and machining the groove of each of the poloidal segments by using the machining strategy. In some embodiments, the step of “calculating a target machining allowance of each of the poloidal segments according to the cross-reconstruction region and a preset segment boundary” is performed through steps of: calculating a total machining allowance based on an area of the cross-reconstruction region; if the preset segment boundary locates in the cross-reconstruction region, calculating an area ratio of two sub-regions divided by the preset segment boundary in the cross-reconstruction region; and calculating the target machining allowance of each of the poloidal segments according to the area ratio and the total machining allowance; and if the preset segment boundary is outside the cross-reconstruction region, distributing the total machining allowance according to a preset distribution strategy to obtain the target machining allowance of each of the poloidal segments. In some embodiments, the step of “generating a machining strategy for each of the poloidal segments according to the target machining allowance and a target groove size” is performed through steps of: calculating a difference between actual machining allowances of adjacent poloidal segments; if the difference is greater than a preset value, correcting the preset segment boundary according to the difference; wherein a corrected preset segment boundary is close to a poloidal segment with a larger actual machining allowance; and generating a groove machining parameter of each of the poloidal segments according to the corrected preset segment boundary and the target groove size; wherein a portion to be cut at an end face of one poloidal segment with a larger actual machining allowance is larger than a portion to be cut at an end face of the other poloidal segment with a smaller actual machining allowance. In some embodiments, the step of “collecting three-dimensional (3D) point cloud data of a surface of individual poloidal segments of a vacuum chamber of a fusion reactor” is performed through steps of: taking a coordinate system of a laser tracker as a first 3D coordinate system; and collecting a 3D point cloud data of a side face of each of the poloidal segments under the first 3D coordinate system; wherein the surface of individual poloidal segments is composed of an end face and the side face; collecting a 3D point cloud data of the end face of each of the poloidal segments by using a measuring arm; and converting the 3D point cloud data of the end face of each of the poloidal segments to the first 3D coordinate system according to a common reference point calibrated between the measuring arm and the laser tracker. In some embodiments, the step of “performing reverse model reconstruction, based on the three-dimensional point cloud data, to generate an actual 3D model of the vacuum chamber of the fusion reactor” is performed through steps of: establishing a second 3D coordinate system in a design software; and generating an ideal 3D model of the vacuum chamber of the fusion reactor in the design software; performing relationship matching between the 3D point cloud data of the surface of each of the poloidal segments and the ideal 3D model until a relationship matching result meets a target convergence accuracy to obtain an optimal fitting relationship between the 3D point cloud data of the surface of each of the poloidal segments and the ideal 3D model; and converting the 3D point cloud data of the surface of each of the poloidal segments to the second 3D coordinate system according to the optimal fitting relationship; and performing model reverse reconstruction on each of the poloidal segments under the second 3D coordinate system to generate the actual 3D model of the vacuum chamber of the fusion reactor. In a second aspect, the present disclosure provides a device for machining a groove of poloidal segments of a vacuum chamber of a fusion reactor, comprising: an acquisition module; a reconstruction module; and a processing module; wherein the acquisition module is configured to acquire a 3D point cloud data of a surface of each of the poloidal segments; the reconstruction module is configured to perform model reverse reconstruction based on the 3D point cloud data to generate an actual 3D model of the vacuum chamber of the fusion reactor and obtain a sectional view of the vacuum chamber of the fusion reactor according to the actual 3D model; and the processing module is configured to extract a cross-reconstruction region between adjacent poloidal segments from the sectional view, calculate a target machining allowance of each of the poloidal segments according to the cross-reconstruction region and a preset segment boundary, and generate a machining strategy for each of the poloidal segments according to the target machining allowance and a target groove size to machine the groove of each of the poloidal segments. In some embodiments, the processing module comprises a distribution unit and a generation unit; wherein the distribution unit is configured to calculate a total machining allowance based on an area of the cross-reconstruction region; the distribution unit is also configured to calculate an area ratio of two sub-regions in the cross-reconstruction region divided by the preset segment boundary if the preset segment boundary locates in the cross-reconstruction region, and calculate the target machining allowance of each of the poloidal segments according to the area ratio and the total machining allowance; and the distribution unit is also configured to distribute the total machining allowance according to a preset distribution strategy if the preset segment boundary is outside the cross-reconstruction region to obtain the target machining allowance of each of the poloidal segments; and the generation unit is configured to calculate a difference between actual machining allowances of adjacent poloidal segments; the generation unit is also configured to correct the preset segment boundary according to the difference if the difference is greater than a preset value, wherein a corrected preset segment boundary is close to a poloidal segment with a larger actual machining allowance; the generation unit is also configured to generate a groove machining parameter of each of the poloidal segments according to the corrected preset segment boundary and the target groove size of each of the poloidal segments; wherein a portion to be cut on an end face of one poloidal segment with a larger actual machining allowance is larger than a portion to be cut on an end face of the other poloidal segment with a smaller actual machining allowance. In some embodiments, the acquisition module is also configured to collect a 3D point cloud data of a side face of each of the poloidal segments under a first 3D coordinate system, wherein the first 3D coordinate system is a coordinate system of a laser tracker; the surface of each of the poloidal segments is composed of an end face and the side face; and the reconstruction module is configured to establish a second 3D coordinate system in a design software, generate an ideal 3D model of the vacuum chamber of the fusion reactor in the design software, perform relationship matching between the 3D point cloud data of the surface of each of the poloidal segments and the ideal 3D model until a relationship matching result meets a target convergence accuracy to obtain an optimal fitting relationship between the 3D point cloud data of the surface of each of the poloidal segments and the ideal 3D model, convert the 3D point cloud data of the surface of each of the poloidal segments to the second 3D coordinate system by using the optimal fitting relationship, and perform model reverse reconstruction on each of the poloidal segments under the second 3D coordinate system to generate the actual 3D model of the vacuum chamber of the fusion reactor. In a third aspect, the present disclosure provides an apparatus, comprising: a processor; a memory; and a computer program; wherein the computer program is stored in the memory, and is configured to be executable on the processor; and the processor is configured to execute the computer program to implement the method for machining a groove of poloidal segments of a vacuum chamber of a fusion reactor. In a fourth aspect, the present disclosure provides a computer-readable storage medium, wherein the computer-readable storage medium stores a computer program; and the computer program is configured to be executed by a processor to implement the method for machining a groove of poloidal segments of a vacuum chamber of a fusion reactor. Compared to the prior art, the present disclosure has the following beneficial effects. In the method provided herein, the actual 3D model of the vacuum chamber of the fusion reactor obtained by reverse model reconstruction can contribute to accurately determining the machining allowance of individual poloidal segments to achieve the secondary distribution of the machining allowance. Moreover, the groove can be redesigned according to the machining allowance of the poloidal segment to eliminate the effect of contour error on the machining precision during the manufacturing process of the poloidal segments, which can effectively improve the machining precision of groove welding. Additional aspects and advantages of this application will be provided partially below, some of which will become apparent from the following description or will be known from the implementation of this application. Embodiments of this application will be described in detail below with reference to the accompanying drawings. The identical or similar reference numerals represent identical or similar elements or elements having the same or similar functions. The embodiments described below are merely illustrative and intended to explain this application, and should not be understood as a limitation to this application. The Tokamak device is considered as the most basic and promising toroidal plasma magnetic confinement nuclear fusion device. The Tokamak device includes a magnetic system, a vacuum chamber, a vacuum system, a blanket system, a divertor system, a power system, a refrigeration system, and a diagnostic system. The vacuum chamber is the heart of the Tokamak device, and the design, material, manufacturing, assembling, sealing and the required ultra-vacuum condition of the vacuum chamber all directly affect the discharge quality of the plasma. Currently, the vacuum chamber of the existing nuclear fusion reactor usually has a double-shell structure with a D-shaped section. For example, the D-shaped section is 15 m in height; the maximum circumferential diameter of the equatorial plane is 25.5 m, and the minimum circumferential diameter is 7.2 m; the shell is made of ultra-low carbon stainless steel material with a thickness of 50 mm and has a total weight of 5600 tons. Given the transportation restrictions and modular assembly, the vacuum chamber is commonly divided into eight 45° sector regions along a toroidal direction, and after fabricated, the eight 45° sector regions are transported to the final assembly site to be assembled into a 360° toroidal vacuum chamber. The existing technical requirements for the ⅛ sector region are mainly described as follows. (1) All main welds shall be fully penetrated, and the weld defect grade shall meet the requirements of ISO 5817:2003B (arc welding) or GB/T 22085.2-2008 (electron beam welding). (2) All welds shall pass 100% nondestructive examination. (3) The tolerance of the overall size should be controlled within ±8 mm. The surface roughness of inner and outer shells is Ra3.2. The leakage rate should be no more than 1.0*10−8 Pa·m·s−1. These technical requirements pose great technical challenges for the shell forming, sub-component manufacturing, welding deformation control and assembly. The ⅛ sector region is divided into a series of functional modules from bottom to top, where the functional modules are relatively independent, universal, and interchangeable. The ⅛ sector region is formed through selection and combination of the functional modules. The ⅛ sector region is composed of two symmetric 1/16 sector regions, and the 1/16 sector region is composed of two symmetric 1/32 sector regions. The 1/32 sector region is divided into four poloidal segments along a poloidal direction. According to the tolerance allocation, the contour deviation of the 1/16 sector region is ±6 mm; the contour deviation of the 1/32 sector region is ±4 mm; and the contour deviation of the poloidal segments is ±3 mm. Before the four poloidal segments are assembled and welded into the 1/32 sector region, the welding grooves need to be machined with a machining precision superior to 0.5 mm. However, there are still the following technical difficulties in the groove machining of the poloidal segment in the prior art. (1) In the manufacturing process, the accumulation of forming errors, assembly errors, and welding deformation will cause the contour deviation of the poloidal segment, which will further affect the machining precision of welding grooves. (2) The machining allowance of the end face of poloidal segment cannot be distributed unevenly, and the machining allowances of some end faces are negative, which requires a secondary allocation. A method, device, and apparatus for machining a groove of a poloidal segment of a vacuum chamber of a fusion reactor, and a computer-readable storage medium will be described with reference to the drawings below. Concerning that the machining allowance of the poloidal segment cannot be accurately determined during machining, which greatly reduces the machining precision of groove welding, this application provides a method for machining a groove of a poloidal segment of a vacuum chamber in a nuclear fusion reactor. In the method provided herein, the actual 3D model of the vacuum chamber of the fusion reactor obtained by reverse model reconstruction can contribute to accurately determining the machining allowance of individual poloidal segments to achieve the secondary distribution of the machining allowance. Moreover, the groove can be redesigned according to the machining allowance of the poloidal segment to eliminate the effect of contour errors on the machining precision during the manufacturing process of poloidal segments, which can effectively improve the machining precision of groove welding. FIG. 1 is a flow chart of a method for machining grooves of poloidal segments of a vacuum chamber of a nuclear fusion reactor according to an embodiment of this application. As shown in FIG. 1, the method includes the following steps. (S101) A 3D point cloud data of a surface of individual poloidal segments of the vacuum chamber of the fusion reactor are collected. The surfaces of each of the poloidal segments is composed of end faces and side faces. As shown in FIG. 2, four poloidal segments of a 1/32 sector region are illustrated. The butt surfaces between adjacent poloidal segments are the end faces, and the other faces are the side faces. In this embodiment, the step (S101) includes the following steps. A coordinate system of a laser tracker is taken as a first 3D coordinate system, and a 3D point cloud data of a side face of each of the poloidal segments is collected under the first 3D coordinate system. A 3D point cloud data of the end face of each of the poloidal segments is collected by using a measuring arm and are converted to the first 3D coordinate system according to a common reference point calibrated between the measuring arm and the laser tracker. The first 3D coordinate system is the real space coordinate system of the laser tracker. It should be understood that in this embodiment, a set of common reference points are measured, combined with the laser tracker and a measuring arm with flexible joint coordinates to collect the 3D point cloud data on the surface of each of the poloidal segments. The laser tracker can collect points evenly due to its wide measurement range, and the measuring arm with the flexible joint coordinates can obtain the 3D point cloud data of the end face of each of the poloidal segments by laser scanning, which can effectively improve the convenience and efficiency of data collection. In the process of data collection, any points of the laser tracker is the center to establish the first 3D coordinate system. The point cloud data obtained by the measuring arm with the flexible joint coordinates is converted to the first 3D coordinate system through common reference point conversion to obtain the 3D point cloud data of the surface of each of the poloidal segments under the same 3D coordinate system, as shown in FIG. 3. (S102) A reverse model reconstruction is performed to generate an actual 3D model of the vacuum chamber of the fusion reactor and acquire a sectional view of the vacuum chamber of the fusion reactor according to the actual 3D model. In this embodiment, the reverse model reconstruction can be performed through many methods, such as a method based on a CATIA V5R20 software reverse engineering module, which is not limited specifically. It can be understood that the actual 3D model of the vacuum chamber of the fusion reactor can be obtained by reverse model reconstruction, and the sectional view of the vacuum chamber of the fusion reactor can be obtained by the sectional view of the actual 3D model, where the sectional view includes a local sectional view, through which the two-dimensional diagram shown in FIG. 4 can be generated. In this embodiment, the step “A reverse model reconstruction is performed to generate an actual 3D model of the vacuum chamber of the fusion reactor” is performed though the following steps. A second 3D coordinate system is established in a design software, and an ideal 3D model of the vacuum chamber of the fusion reactor is generated in the design software. Relationship matching is performed between the 3D point cloud data of the surface of each of the poloidal segments and the ideal 3D model until a relationship matching result meets a target convergence accuracy to obtain an optimal fitting relationship between the 3D point cloud data of the surface of each of the poloidal segments and the ideal 3D model. The 3D point cloud data of the surface of each of the poloidal segments are converted to the second 3D coordinate system according to the optimal fitting relationship. The reverse model reconstruction is performed on each of the poloidal segments under the second 3D coordinate system to generate the actual 3D model of the vacuum chamber of the fusion reactor. The second 3D coordinate system is a virtual coordinate system established in the design software. The design software can be selected according to actual needs, such as the Spatial Analyzer software. The ideal 3D model refers to the design model obtained by using the design software, such as the computer-aided design (CAD) model. The step (102) is performed specifically as follows. (1) By using the design software, and based on the principle of least square, the optimal fitting is performed on the contour data of each of the poloidal segments and the ideal 3D model of each of the poloidal segments to solve the space conversion relationship, where the optimal fitting refers to the relationship matching between the measured data and the designed model to obtain the optimal solution, and the relationship matching is performed until the result meets a certain convergence accuracy. (2) According to the space conversion solved by the optimal fitting, the data measured by the laser tracker and the measuring arm with the flexible joint coordinates can be converted from the first 3D coordinate system to the second 3D coordinate system. The first 3D coordinate system can be consistent with the second 3D coordinate system to realize the unification of the collected data of each of the poloidal segments. For instance, as shown in FIG. 5, the collected data of the poloidal segments illustrated in FIG. 2 and FIG. 3 can be unified. The space conversion can be a space conversion matrix, which can be applied to the whole data. (3) Under the second 3D coordinate system, the reverse model reconstruction is performed, and the 3D point cloud data measured by the measuring arm with the flexible joint coordinates and the laser tracker is processed to generate the actual 3D model. In this embodiment, as shown in FIG. 6, the step of “Under the second 3D coordinate system, the reverse model reconstruction is performed, and the 3D point cloud data measured by the measuring arm with the flexible joint coordinates and the laser tracker is processed to generate the actual 3D model” is performed through the following steps. The 3D point cloud data is processed with denoising and filtering to create a triangular network. Scanning curves and B-spline curves are created in the triangular network to generate a curvilinear network. A bridge curved surface is reconstructed in the curvilinear network to generate a smooth curved surface. The smooth curved surface is processed with a closed curved surface to generate the actual 3D model. (S103) A cross-reconstruction region between two adjacent poloidal segments is extracted from a sectional view. A target machining allowance of each of the poloidal segments is calculated according to the cross-reconstruction region and a preset segment boundary. A machining strategy for a groove of each of the poloidal segments is generated according to the target machining allowance and a target groove size s. The groove of each of the poloidal segments is machined by using the machining strategy. The preset segment boundary refers to the segment boundary between adjacent poloidal segments in the ideal 3D model under the second 3D coordinate system. It should be understood that in this embodiment, the deviation between the actual 3D model and the ideal 3D model can be compared based on reverse engineering technology to complete secondary distribution of the machining allowance of each of the poloidal segments in the vacuum chamber and design grooves on the actual 3D model. For example, grooves can be designed on the actual 3D model to output a machining model, then the grooves are machined according to the machining model by using a machine tool, which can eliminate the contour errors of the actual manufacturing poloidal segments and ensure the machining precision of groove welding. The machining of the grooves can use a floor-type boring and milling machine, gantry milling, and other machining methods. In this embodiment, the target machining allowance of each of the poloidal segments is calculated as follows. A total machining allowance is calculated based on an area of the cross-reconstruction region. If the preset segment boundary locates in the cross-reconstruction region, an area ratio of two sub-regions divided by the preset segment boundary in the cross-reconstruction area is calculated, and the target machining allowance of each of the poloidal segments according to the area ratio and the total machining allowances is calculated. If the preset segment boundary is outside the cross-reconstruction region, the total machining allowance is distributed according to a preset distribution strategy to obtain the target machining allowance of each of the poloidal segments. The larger the area of the cross-reconstruction region is, the larger the machining allowance exists. The actual machining allowance can be determined according to the corresponding relationship between the area of the cross-reconstruction region and the machining allowance. The preset distribution strategy can be set according to the actual situation, which is not limited specifically. For example, a preset distribution strategy is described as follows. The distribution ratio of one of two adjacent poloidal segments near the preset segment boundary is greater than that of the other poloidal segment. As shown in FIG. 4, the preset segment boundary is located above the cross-reconstruction region and near the poloidal segment 2, therefore the machining allowance of the poloidal segment 2 is greater than that of the poloidal segment 3, for instance, the machining allowance distributed to the poloidal segment 2 can be 80% and that to the poloidal segment 3 can be 20%. As shown in FIG. 4, as the four poloidal segments are all under the same coordinate system, after the reverse model reconstructions of the end face of each of the poloidal segments is completed, the actual 3D model and the ideal 3D model of each end face can be compared and analyzed to compare the machining allowance of the actual 3D model and the ideal 3D model of each of the poloidal segments. In this embodiment, the step of “A machining strategy for a groove of each of the poloidal segments is generated according to the target machining allowance and a target groove sizes.” is performed through the following steps. A difference between actual machining allowances of adjacent poloidal segments is calculated. If the difference is greater than a preset value, the preset segment boundary is corrected according to the difference, where a corrected preset segment boundary is close to a poloidal segment with a larger actual machining allowance. A groove machining parameter of each of the poloidal segments is generated according to the corrected preset segment boundary and the target groove size, where a portion to be cut at an end face of one poloidal segment with a larger actual machining allowance is larger than a portion to be cut at an end face of the other poloidal segment with a smaller actual machining allowance. The preset value can be set or calibrated according to the actual situation, which is not limited specifically. It should be understood that in this embodiment, through the cross-reconstruction region and the comparison of the preset segment boundary, the machining allowance and the adjustment of the preset segment boundary can be determined to improve the accuracy of groove machining. In this embodiment, the actual machining allowance between adjacent poloidal segments determines whether the preset segment boundary needs to be corrected. For example, as shown in FIG. 4, as the machining allowance of an inner shell in poloidal segment 2 is large and that in the poloidal segment 3 is negative, the preset segment boundary can be shifted toward the poloidal segment 2. As a result, the correction of the preset segment boundary can promote the accuracy of groove machining. After the secondary distribution of the machining allowance, the groove is designed based on the actual 3D model to output a model. The groove of each of the poloidal segments can be machined via a machining tool based on the model with a machining precision of better than 0.5 mm. In the method, the actual 3D model of the vacuum chamber of the fusion reactor obtained by reverse model reconstruction can contribute to accurately determining the machining allowance of individual poloidal segments to achieve the secondary distribution of the machining allowance. Moreover, the groove can be redesigned according to the machining allowance of the poloidal segment to eliminate the effect of contour error on the machining precision during the manufacturing process of the poloidal segments, which can effectively improve the machining precision of groove welding. A device for machining a groove of poloidal segments of a vacuum chamber of a fusion reactor is provided. As shown in FIG. 7, the device 10 includes an acquisition module 100, a reconstruction module 200, and a processing module 300. The acquisition module 100 is configured to acquire a 3D point cloud data of a surface of each of the poloidal segments. The reconstruction module 200 is configured to perform reverse model reconstruction based on the 3D point cloud data to generate the actual 3D model of the vacuum chamber of the fusion reactor and obtain a sectional view of the vacuum chamber of the fusion reactor according to the actual 3D model. The processing module 300 is configured to extract the cross-reconstruction region between adjacent poloidal segments from the sectional view, calculate a target machining allowance of each of the poloidal segments according to the cross-reconstruction region and a preset segment boundary, and generate a machining strategy for each of the poloidal segments according to the target machining allowance and a target groove size s to machine the groove of each of the poloidal segments. In some embodiments, the processing module 300 includes a distribution unit and a generation unit. The distribution unit is configured to calculate a total machining allowance based on an area of the cross-reconstruction region. The distribution unit is also configured to calculate an area ratio of two sub-regions in the cross-reconstruction region divided by the preset segment boundary if the preset segmenting boundary falls in the cross-reconstruction region, and the target machining allowance of each of the poloidal segments is calculated according to the area ratio and the total machining allowance. If the preset segment boundary is outside the cross-reconstruction region to obtain the target machining allowance of each of the poloidal segments. The generation unit is configured to calculate a difference between actual machining allowances of adjacent poloidal segments. The generation unit is also configured to correct the preset segment boundary according to the difference if the difference is greater than a preset value, where a corrected preset segment boundary is close to a poloidal segment with a larger actual machining allowance. the generation unit is also configured to generate a groove machining parameter of each of the poloidal segments according to the corrected preset segment boundary and the target groove size of each of the poloidal segments, where a portion to be cut on an end face of one poloidal segment with a larger actual machining allowance is larger than a portion to be cut on an end face of the other poloidal segment with a smaller actual machining allowance. In some embodiments, the acquisition module is also configured to collect a 3D point cloud data of a side face of each of the poloidal segments under a first 3D coordinate system, where the first 3D coordinate system is a coordinate system of a laser tracker; the surface of each of the poloidal segments is composed of an end face and the side face. The acquisition module is also configured to collect a 3D point cloud data of the end face of each of the poloidal segments by using a measuring arm, and convert the 3D point cloud data of the end face of each of the poloidal segments to the first 3D coordinate system according to a common reference point calibrated between the measuring arm and the laser tracker. In some embodiments, the reconstruction module 200 is configured to establish a second 3D coordinate system in a design software, generate an ideal 3D model of the vacuum chamber of the fusion reactor in the design software, perform relationship matching between the 3D point cloud data of the surface of each of the poloidal segments and the ideal 3D model until a relationship matching result meets a target convergence accuracy to obtain an optimal fitting relationship, convert the 3D point cloud data of the surface of each of the poloidal segments to the second 3D coordinate system by using the optimal fitting relationship, and perform reverse model reconstruction on each of the poloidal segments in the second 3D coordinate system to generate the actual 3D model of the vacuum chamber of the fusion reactor. It should be noted that the above explanation of the method is also applicable to the device provided herein, which will not be repeated here. In the device provided herein, the actual 3D model of the vacuum chamber of the fusion reactor obtained by reverse model reconstruction can contribute to accurately determining the machining allowance of individual poloidal segments to achieve the secondary distribution of the machining allowance. Moreover, the groove can be redesigned according to the machining allowance of the poloidal segment to eliminate the effect of contour error on the machining precision during the manufacturing process of the poloidal segments, which can effectively improve the machining precision of groove welding. Referring to FIG. 8, an apparatus for machining a groove of poloidal segments of a vacuum chamber of a fusion reactor is provided, which includes a memory 1201, a processor 1202, and a computer program, where the computer program is stored in the memory 1201 and is configured to be executable on the processor 1202; and the processor 1202 is configured to execute the computer program to implement the method for machining a groove of the poloidal segments of a vacuum chamber of a fusion reactor. In some embodiments, the apparatus further includes a communication interface 1203, where the communication interface 1203 is configured to communicate the memory 1201 with the processor 1202; and the memory 1201 may include a fast random-access memory (RAM), or a non-volatile memory, such as at least one disk memory. If the memory 1201, the processor 1202, and the communication interface 1203 are independent, the communication interface 1203, the memory 1201, and the processor 1202 can be connected and communicated through a bus. The bus can be an industry-standard architecture (ISA) bus, a peripheral component (PCI) bus, or an extended industry standard architecture (EISA) bus. The bus can be classified into an address bus, data bus, control bus, and so on. For ease of representation, a single thick line is configured to illustrate the bus, as shown in FIG. 12, but it does not mean that there is only one bus or one type of bus. In some embodiments, if the memory 1201, the processor 1202, and the communication interface 1203 are integrated on a chip, the memory 1201, the processor 1202, and the communication interface 1203 are communicated through an internal interface. The processor 1202 can be a central processing unit (CPU), or an application-specific integrated circuit (ASIC), or one or more integrated circuits configured to implement the embodiments of this application. A computer-readable storage medium is also provided herein, where the computer-readable storage medium stores a computer program; and the computer program is configured to be executed by a processor to implement the method for machining a groove of poloidal segments of a vacuum chamber of a fusion reactors as described above. As used herein, the reference terms “one embodiment”, “embodiments”, “examples”, “specific examples”, and “some examples” mean that the specific features, structures, materials, or characteristics described herein are contained in at least one embodiment or example of this application. The indicative representations of the above terms are not necessarily specific to the same embodiments or examples. Furthermore, the specific features, structures, materials or characteristics described herein can be combined in any one or N embodiments or examples in an appropriate manner. In addition, without contradiction, different embodiments or examples and the characteristics of the different embodiments or examples can be combined by those skilled in the art. Furthermore, the terms “first” and “second” are only intended to describe and do not necessarily require or imply that there is any relative importance or the number of technical features. Thus, features that are limited with the term “first” or “second” may include, explicitly or implicitly, at least one of these features. The term “N” means at least two (such as two, three, etc.), unless otherwise expressly specified. Flow charts or any process or method description described in other ways include one or N modules, fragments, or parts of the codes of the executable instructions configured to implement a custom logic function or process steps. Moreover, it should be understood by those skilled in the art that the preferred modes of the implementation can perform functions without being in the order shown or discussed, including a simultaneous manner or reverse order based on the functions involved. Logic and/or steps represented in the flow charts or otherwise described herein, for example, a fixed list of executable instructions that can be considered to implement logical functions, can be implemented in any computer-readable medium for use by instruction execution systems, devices, or equipment (such as computer-based systems, including processor systems, or other systems that can read and execute instructions from, or other devices in combination with these instruction execution systems, devices, or equipment. As used herein, computer-readable media can be any device that may contain store, communicate, propagate, or transmit programs for the execution of a system, device, or equipment, or other devices in combination with these instruction execution systems, devices, or equipment. computer-readable media include electrical connecting part (electronic device) with one or N cabling, laptop enclosure (magnetic device), random access memory (RAM), read-only memory (ROM), erasable read-only memory (EPROM or flash memory), fiber optic device, and portable compact disc read-only memory (CDROM). Furthermore, computer-readable media can even be paper or other suitable media on which the program can be printed. For example, by optical scanning on paper or other media, followed by editing, interpretation, or other appropriate processing to obtain the program electronically and then store it in a memory. It should be understood that the various parts of this application can be implemented with hardware, software, firmware, or a combination thereof. In the above embodiments, N steps or methods can be implemented using software or firmware stored in a memory and executed by the appropriate instruction execution system. If N steps or methods are be implemented using a hardware, it can be achieved by a discrete logic circuit with logic gate circuits that can endow data signals logic gate function, a special integrated circuit with suitable combination logic gate circuits, a programmable door array (PGA), a field-programmable door array (FPGA), or a combination thereof. It should be understood by those skilled in the art that all or part of the steps disclosed in the above embodiments can be performed by a program to instruct the relevant hardware. The program can be stored in a computer-readable storage medium, the program includes one of the steps or a combination thereof. Furthermore, functional units in the embodiments of this application can be integrated into a processing module, or are independent, or two or more units can be integrated into a single module. The above-mentioned integrated modules can be implemented either in the form of hardware or in the form of software function modules. The integrated modules can also be stored in a computer-readable storage medium if they are implemented in the form of software function modules and sold or used as an independent product. The above-mentioned storage media can be a read-only memory, a disk, or a compact disc. Although the embodiments of the disclosure have disclosed and described above, it should be understood that these embodiments are merely exemplary and are not intended to limit this application. Various modifications, changes, replacements, and variations made by those skilled in the art without departing from the scope of the disclosure should fall within the scope of the disclosure defined by the appended claims. |
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abstract | An x-ray imaging system includes a detector positioned to receive x-rays, and an x-ray tube coupled to a mount structure. The x-ray tube is configured to generate x-rays toward the detector and includes a target, a cathode cup, an emitter attached to the cathode cup and configured to emit a beam of electrons toward the target, the emitter having a length and a width, and a one-dimensional grid positioned between the emitter and the target and attached to the cathode cup at one or more attachment points. The one-dimensional grid includes a plurality of rungs that each extend in a direction of the width of the emitter, and the plurality of rungs are configured to expand and contract relative to the one or more attachment points without substantial distortion with respect to the emitter. |
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abstract | An atomic beam source includes a tubular cathode that includes an emission portion that includes an emission port through which an atomic beam can be emitted, a rod-shaped first anode disposed inside the cathode, and a rod-shaped second anode disposed inside the cathode and spaced from the first anode. At least one selected from the group consisting of a shape of the cathode, a shape of the first anode, a shape of the second anode, and a positional relationship between the cathode, the first anode, and the second anode is predetermined so that emission of sputter particles resulting from collision of cations, which have been generated by plasma between the first anode and the second anode, with at least one selected from the cathode, the first anode, and the second anode is reduced. |
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050135228 | summary | BACKGROUND AND SUMMARY OF THE INVENTION Nuclear power plants and other heat-generating power plants such as coal, oil, and gas power plants, are normally equipped with a circulation system for water, in which steam which is generated in a reactor vessel or in a steam generator connected to the circulation system is passed to a steam turbine, from there to a condenser and condensate formed therein, after cleaning and preheating, is returned to the reactor vessel of the steam generator. In operation of power plants of the above-mentioned kinds different corrosion products are formed in the circulation system, inter alia in the form of oxides containing one or more metals of the kinds included in construction materials in the circulation system such as, above all, iron and further, inter alia, nickel, cobalt, chromium, manganese, titanium, molybdenum, zinc and zirconium. The corrosion products occur both as ionic dissolved compounds and as particulate undissolved compounds. In those cases where a nuclear reactor vessel is part of the circulation system, the corrosion products will become radioactive. Besides being influenced by the temperature, the amount and the nature of the corrosion products are influenced especially by the pH value of the water and the concentration of hydrogen and oxygen dissolved in the water. The corrosion products are therefore indicators of important properties of the water. By continuously investigating the corrosion products, it can be determined whether changes in the corrosion environment have occurred which require action regarding the operation of the power plant to eliminate the cause of the changes, such as, inter alia, changes of the pH of the water and, if a nuclear reactor vessel is included in the circulation system, inter alia the degree of radioactive contamination in the circulation system and the degree of oxide growth of cladding material for the fuel and of box material. It is, of course, of decisive importance for the operation of a power plant of the above-mentioned kinds that the water in the circulation system is as clean as possible to counteract the occurrence of corrosive coatings and an ensuing deteriorated economy. When it comes to determining the chemical composition and the content of ionic dissolved compounds in water in water circulation systems of the kind described, automatic sampling systems have been produced, with the aid of which the nature and content of ionic compounds can be continuously followed during operation of the power plant. For analysis of particulate compounds, no corresponding automatic technique has been available. It has been necessary to separate the particulate compound from samples taken of the water on filters and then to determine the compounds by conventional methods by measuring directly on the filter or by dissolving the compounds in a solvent in a separate operation and analyzing the solution thus obtained in another separate operation. For analysis of this solution, a great number of methods can be used. In those cases where the analysis sensitivity is too low, the contents in the solution can be increased by using for analyses a larger water sample and hence a longer filtering time (enrichment time). Determination methods for solid samples for direct measurement of the particulate compounds on the filter are not applicable to water in water circulation systems of the kind described because of insufficient sensitivity. According to the present invention, particulate compounds can be determined automatically in a sampling system in a reliable and simple manner, which includes a filtering of samples from the water in water circulation systems of the above-described kind or of liquid flows of another kind as well as dissolution of the particulate compounds, collected on the filter, in a solvent and determination of the nature and content of one or more of the particulate compounds in the solution obtained. For the dissolution known solvents for the particulate particles in question are used, and for analysis of the solution known analysis methods are used. The invention is applicable not only to the analysis of particulate compounds in water in water circulation systems in power plants of the kinds described in the introduction, but also to the determination of particulate compounds in water of another kind and in liquids of another kind, such as spirit solutions and solutions of organic solvents, to the extent that these contain particulate compounds. The present invention relates to a method of determining the chemical composition and the content of one or several particulate compounds in a flow of a liquid, whereby samples of the liquid are taken out from the flow and particulate compounds in the sample taken are collected on a filter through which the liquid in the sample is caused to pass, whereafter the collected particulate compounds are dissolved in a solvent while forming a solution in which the chemical composition and the amount of one or more of the particulate compounds are determined, characterized in that the filter is arranged in a closable container of a material permeable to microwaves; that samples of a fixed size of the liquid in the flow are supplied batchwise to the container on one side of the filter via an openable and a closable connection between the liquid flow and the container; that liquid, having passed through the filter, from each liquid sample supplied batch-wise is discharged from the container via an openable and a closable outlet on the other side of the filter; that after closing of the connection between the liquid flow and the container and of the outlet for the liquid having passed through the filter, a fixed amount of the solvent for the particulate compounds is supplied via an openable and a closable inlet to the container with particulate compounds from each liquid sample, supplied batchwise, collected on the filter and with the liquid from the liquid sample, supplied batchwise, discharged from the container; that the collected particulate compounds from each liquid sample, supplied batchwise, are dissolved in the solvent while heating the contents of the container by a microwave field from a microwave oven in which the container is arranged; and that the solution of the particulate compounds, thus obtained, in each liquid sample supplied batchwise is discharged via an openable and a closable outlet to an analysis apparatus, in which the composition and the content of one or more of the particulate compounds are determined. The filter consists of a material which is resistant to the liquid to be analyzed and to the used solvent. Among suitable materials may be mentioned polytetrafluoroethylene and other fluorinated hydrocarbon polymers, for example polychlorotrifluoroethylene, as well as other resistant resins; further, inter alia, porcelain filters and sintered resistant particulate materials, for example sintered glass filters. The hole size of the filter is chosen in view of the size of the particulate compounds in the liquid flow. In most cases a hole size within the interval 0.2 .mu.m-1.0 .mu.m is suitable. The container of a material permeable to microwaves also consists of a material which is resistant to the liquid to be analyzed and to the used solvent. Among suitable materials may be mentioned polytetrafluoroethylene and other fluorinated hydrocarbon polymers, for example polychlorotrifluoroethylene and glass. The volume of the container may suitably amount to 5-50 ml. The particulate compounds are collected on the filter for a certain period of time. The time is chosen, depending on the particle content, such that a suitable amount of particles are collected on the filter. This amount should be such that it is suitable for analysis of the solution of the dissolved particulate compounds. The solvents which are used for dissolution of the particulate compounds collected on the filter are, of course, dependent on the nature of the particulate compounds. For dissolution of corrosion products in water in a water circulation system in a nuclear power plant, as well as in other power plants of the kind described above, mixtures of acids may be used, preferably oxidizing such mixtures, as well as formers of chelate complexes. Mixtures of acids can be used for decomposition of most types of corrosion products of the kind described, whereas chelate formers are more specific. The chelate solutions usually have an optimum pH for dissolution in the interval 2-8. As examples of useful mixtures of acids may be mentioned mixtures of at least two of the acids hydrochloric acid, nitric acid, and perchloric acid. As examples of useful formers of chelate complexes may be mentioned ethylenediaminotetra acetic acid, citric acid and thioglycolic acid (HS--CH.sub.2 --COOH). The latter is particularly effective in mixtures with hydrochloric acid. The time for dissolving a sample of particulate compounds varies with the sample quantity, the sample composition, the sample structure, the degree of atomization, the temperature, and the strength, pH and composition of the mixture of acids or the chelate solution. By heating the sample by microwaves, the rate of dissolution may be greatly reduced, from the order of magnitude of hours without microwave heating to the order of magnitude of minutes with microwave heating. The increased rate is assumed to be due to an increased convection in the solution and to an absorption of energy directly and locally in the grains of the particulate material. The analysis of the solution of the particulate compounds is performed with known methods, for example by ion chromatography or by atom absorption spectrometry. The method according to the invention operates on line, i.e., the sampling equipment used is connected to the liquid flow in which the nature and content of the particulate compounds are to be analyzed. All measures, such as opening and closing of valves, heating processes and times for different operations, are controlled by automatic control, preferably by computer control, for example comprising the use of microprocessors. In this way, a series of results is obtained from each sample taken batchwise from the liquid flow, thus obtaining an automatic determination of the particualte compounds for the whole of the time during which the determination is carried out. |
abstract | A winder includes an attachment to an external portion of a glove port fitted on a closed vacuum chamber, to roll a glove or a bag carried by the glove port outside the chamber. |
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abstract | A electronic blackbody cavity is provided. The electronic blackbody cavity comprises an inner surface; a chamber surrounded by the inner surface; an opening configured to make an electron beam enter the chamber; and a porous carbon material layer located on the inner surface. The porous carbon material layer consists of a plurality of carbon material particles and a plurality of micro gaps. The plurality of micro gaps are defined by the plurality of carbon material particles. A secondary electron detection device using the electronic blackbody cavity is also provided. |
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053696778 | abstract | A testing device for load-testing of specimens (3) in a nuclear reactor environment is attached to one of the pipes (1) of the reactor for conveying a first medium under pressure and provides a first space (14) in open communication with the pipe (1). In the first space (14) a movable pull rod (15) is arranged, one end of which is attached to one half (16) of a specimen (3) arranged in the space (14). The other end of the pull rod (15) is joined to a tensile force device capable of being influenced by the first medium for achieving a tensile stress in the specimen (3). |
044184220 | abstract | A device for positioning a holder for an aiming sleeve for implanted bone nail fasteners with respect to a Roentgen ray source includes a support (25) for supporting the holder (7). A base plate (12) and a pivotable bracket plate (3) receives the support means and is mountable on the housing of the Roentgen ray source for spacedly positioning the holder from the source. The support is formed as a fork-like element (26, 28, 29) retained in sleeves (38, 39) on the bracket plate by means of locking cams (8). |
045004870 | description | DETAILED DESCRIPTION Various reactor accidents can cause pressure surges in the reactor coolant. When the reactor coolant is a material such as liquid sodium it is desired to attenuate the pressure surge without violation of the normal coolant boundary to prevent chemical interactions such as sodium burning. Conventional systems, using blow-out diaphragms, are therefore not optimum. FIRST EMBODIMENT The first embodiment of this invention is applicable to liquid filled pipes, especially coolant filled pipes of nuclear reactor systems. Refer to FIG. 1. A pipe 1 containing coolant 2 has a section of its length being of reduced thickness as compared to the pipe wall thickness, this section 3 moreover optionally having convolutions called flutes 4, visible in FIG. 2, herein. Fluted section 3 is surrounded by a container 5, with the space between container 5 and fluted section 3 being filled with crushable metal foam 6, which is shaped appropriately to fit within the flutes. Metal foam is a material that exhibits controlled energy absorption properties which can be tailored to meet specific requirements. A variety of characteristics of the foam, such as metal alloy composition or void size, may be altered to change the specific energy absorption which occurs during crushing of the foam. Metal foam can be machined to fit desired geometric configurations such as the fluted section 3 of this embodiment. Metal foam is available commercially from several sources, three being (1) Energy Research and Generation, Inc., of Oakland, Calif., (2) Foametal Division of Hogen Industries Inc. of Willoughby, Ohio and (3) Astro Met Associates Inc., of Cincinatti, Ohio. Metal foam is described in detail by the following United States patents which are incorporated herein by reference: U.S. Pat. No. 4,099,961 to Patten, U.S. Pat. No. 3,834,881 to Niebylski, and U.S. Pat. No. 2,553,016 to Sosnick. The operation of the first embodiment during a pressure surge will now be described. The pressure surge in coolant 2 in FIG. 1 is presumed to arrive as a pressure wave at end 7 of pipe 1, moving toward end 8. The surge is moreover presumed to be of such magnitude as to be hazardous to critical structures such that the fluted section 3 strength has been planned to allow the following response to the wave. Fluted section 3 expands radially outward from the center of pipe 1, the effect of the flutes being to allow such expansion without rupture. During expansion of fluted section 3, metal foam 6 is crushed, thereby absorbing energy and attenuating the pressure surge. The length L of the fluted section 3 must be sufficient that the pressure wave is reduced in intensity to a safe level prior to passing beyond section 3. The overall design of the system is planned to fully attenuate expected (or maximum possible) pressure surges prior to expansion of the fluted section to the point where further expansion must occur with plastic deformation of the flute material with consequent risk of rupture and coolant spillage. Metal foam 6 may also serve as a radiation shield and thermal insulator. Since the foam 6 is crushed during attenuation of a pressure surge, and the fluted section is deformed, the unit must be replaced when the fluted section is just fully expanded, or earlier. It is not presently envisioned that resilient or recoverable flutes and foam will be available. An estimate of the potential energy absorbing capability of the first embodiment will now be given, with reference to FIGS. 1 and 2. For a pipe of radius r with N flutes of flute length L, and crushable foam thickness t, it can be shown that: ##EQU1## where: E is the energy absorption per unit length L to completely expand the flute; .sigma. is the yield stress of the foam; and PA1 .epsilon. is the final yield strain of the foam; and, PA1 a is the ratio of the maximum to average radius r of the fluted pipe. PA1 .sigma..sub.y is the desired pressure to be attenuated PA1 .epsilon..sub.f is the final foam yield strain, and PA1 V is the foam volume Table I presents a calculated energy absorption for the indicated data. TABLE I ______________________________________ .sigma. 500 PSI .epsilon. 75% r 18 inches t 9.27 inches N 8 flutes a 0.155 E 3.69 .times. 10.sup.5 in-lbs/in for a just fully expanded flute ______________________________________ SECOND EMBODIMENT Refer to FIG. 3. A pad 9 of crushable metal foam 6 is disposed beneath the head structure 10 of a reactor. As a result of a hypothetical accident in a reactor, impact of coolant pool 11 with head structure 10 may occur. Pad 9 will be crushed between plates 12, absorbing energy and attenuating the pressure surge. Foam 6 serves also as a radiation and thermal shield. For crushable foam under-head shielding, potential energy absorption can be defined through EQU E=.sigma..sub.y .epsilon..sub.f V where: Considering, for example, an under-head configuration with: EQU .sigma..sub.y =300 psi EQU .epsilon..sub.f =0.75 EQU V=2.32.times.10.sup.6 in.sup.3 with 48 inches of metal foam, the energy absorbed becomes: ##EQU2## Absorption of such energy reduces head and upper vessel accident loads in a very significant way. THIRD EMBODIMENT Refer to FIG. 4. In this embodiment, crushable metal foam 6 is arranged outside a reactor vessel 13, between said vessel 13 and a plate 14. The functions of the foam are identical to the first two embodiments. |
051981840 | summary | BACKGROUND OF THE INVENTION The present invention relates to a reactor containment vessel containing a reactor pressure vessel, and more particularly, to a reactor pressure vessel supporting pedestal having an improved structure. Generally, as shown in FIG. 7, a reactor primary containment vessel (PCV) 1 contains a reactor pressure vessel (RPV) 2. The PCV 1 is composed of an outer wall structure W, in the inside of which an upper drywell 5 is formed and a lower drywell 7 is also formed below the upper dry well 7. A suppression chamber 8 for absorbing steam energy discharged in the upper and lower dry wells 5 and 7 at a reactor accident is formed inside the wall structure W. In the upper drywell 5, the RPV 2, a line 3 connected to the RPV 2 and an air conditioning system 4 are disposed, and in the lower drywell 7, a control rod driving mechanism 6 and others are disposed. The RPV 2 is supported by a pedestal 12 which is supported at one end by the outer wall structure W through a diaphragm floor 22 and has a cylindrical structure surrounding the RPV 2. A line 9, a cable 10 and a duct 11 are also arranged in the upper and lower drywells 5 and 7, and a connecting vent 13 is formed to the pedestal 12 to pass the line 9, the cable 10 and the duct 11. In an actual design, a plurality of these line, cable and duct may be arranged, but in the illustration, only the cables 10 are shown as plural. This connecting vent 13 serves as a flow passage for guiding the steam discharged from the line, which is broken in an reactor accident, into a vent pipe 14 which is disposed at the lower portion of the pedestal 12. The steam is then guided into the suppression chamber 8 in which the steam is condensed as a suppression pool water 25. In such accident, since non-condensable gas such as nitrogen gas is also flown into the suppression chamber 8 as well as the steam, the pressure inside the suppression chamber 8 increases. Upon breaking the line or duct, the pressure inside the upper and lower drywells 5 and 7 also increase. However, coolant is supplied through a core cooling system at such emergency, and when the broken portion of the line or duct is filled up with the coolant, the steam in the upper and lower dry wells 5 and 7 are condensed as suppression pool water 25, thus rapidly decreasing the pressure inside the drywells 5 and 7. When the pressure inside the drywells 5 and 7 is rapidly lowered, a pressure load is severely applied to the diaphragm floor 22 formed as a partition wall sectioning the upper and lower drywells 5 and 7. In order to prevent such pressure load from appling to the diaphragm floor 22, vacuum breakers 24 are mounted to the pedestal by means of mounting or fixingpipes 23, respectively, as shown in FIGS. 9 and 10. Namely, when a difference in pressures inside the upper drywell 5 and the lower drywell 7 exceeds a predetermined value, the vacuum breakers 24 are operated to flow the gas inside the lower drywell 7 into the suppression chamber 8. These vacuum breakers 24 are disposed in plural numbers along the circumferential direction of the pedestal 12 as shown in FIG. 10. Generally, in the PCV 1, a plurality of internal pumps are arranged inside the PCV 1 along a circumferential direction of a reactor core disposed in the RPV 2 with predetermined spaces with each other and a plurality of vent pipes 14 are also arranged along the circumferential direction of the RPV 2 generally at positions corresponding to the internal pumps. Namely, as shown in FIG. 8, the pedestal 12 is sectioned into a plurality of the connecting vents or passages 13 and a plurality of concrete wall sections 12a, which are alternately arranged along the circumferential direction of the pedestal 12. The pedestal 12 is lined by steel plates, for example, and accordingly, each of the connecting vents 13 is defined by the adjacent concrete sections 12a and the inner and outer steel plates. In such arrangement, however, as shown in FIG. 8 shown as sectional view taken along the lines VIII--VIII in FIG. 7, the line 9, the cables 10, the duct 11 and other elements are disposed in the connecting vent 13 above each of the vent pipe 14, so that it is difficult to ensure a sufficient steam flow area for the vent pipe 14 at a time of accident of the reactor. Moreover, as shown in FIG. 9 or 10, the fixing pipe 23 to which the vacuum breaker 24 is secured penetrates the connecting vent 13 above the vent pipe 14, so that the location of the vacuum breaker 24 and the fixing pipe 23 is complicated, thus being disadvantageous. SUMMARY OF THE INVENTION An object of the present invention is to substantially eliminate defects or drawbacks encountered in the prior art and to provide a reactor containment vessel including a pedestal having an improved structure capable of ensuring an enough space for steam flow in the drywell to a vent pipe at a reactor accident. Another object of the present invention is to provide a reactor containment vessel including a pedestal having an improved structure to which a vacuum breaker means is mounted at an appropriate position for easy construction. These and other objects can be achieved according to the present invention by providing a reactor containment vessel having an outer wall structure an inside of which is divided into upper and lower drywells by means of a diaphragm floor and in which a suppression chamber is arranged, a reactor pressure vessel is supported by a pedestal and a line, a cable and a duct are disposed in and between the upper and lower drywells, the pedestal having a cylindrical structure surrounding the reactor pressure vessel, the pedestal comprising a plurality of concrete wall sections and a plurality of connecting vent sections which are arranged alternately along a circumferential direction of the cylindrical pedestal, wherein the line, the cable and the duct are arranged in each of the connecting vent sections and a vent pipe is arranged in each of the concrete sections so that the line, the cable and the duct and the vent pipe occupy different positions in their cross sections, the vent pipe having an open end opened to the drywell. In modified or preferred embodiments, a vacuum breaker means is further disposed in the reactor containment vessel at a portion above the open end of the vent pipe, the vacuum breaker means including a fixing pipe for mounting a vacuum breaker to the pedestal, the fixing pipe having one end opened to the drywell. The vent pipe may have an end portion extended to the diaphragm floor and opened to the drywell and a vacuum breaker means is further disposed in the reactor containment vessel, the vacuum breaker means being mounted to the extended portion of the vent pipe and including a fixing pipe for mounting a vacuum breaker to the pedestal, the fixing pipe having one end opened to an inside of the vent pipe. According to the characters and structures of the reactor containment vessel of the present invention, the structure of the pedestal is sectioned into a plurality of connecting vent sections in each of which the line, the cable and the duct are arranged and a plurality of concrete wall sections in each of which the vent pipe is disposed. Accordingly, the portion of the drywell above the open end of the vent pipe is formed as a vacant space in which no cable and line are arranged, whereby the steam flow is not obstructed at the reactor accident. In this connection, the vacuum breaker is disposed easily with a simple manner in association with the arrangement of the vent pipe, whereby the pedestal having an improved structure can be provided. The nature and further characteristic features of the present invention will be made more clear from the following descriptions with reference to the accompanying drawings. |
050531859 | abstract | A material analyzer includes a container; a radiation source disposed within the container; a detector disposed within the container for detecting radiation that is secondarily emitted by a material sample within an activation region located between the radiation source and the detector when the material sample is bombarded by radiation from the radiation source and for producing signals in response to said secondarily emitted radiation; a computer for processing the signals to determine the content of the material sample; and a carousel for transporting a material sample from outside the container to within the activation region. The carousel is disposed about an axis of rotation located between the activation region and a receiving region located outside the container, and has a pair of sample containment chambers that are alternately moved between the receiving region and the activation region when the carousel is rotated about said axis. Material samples are placed in buckets for loading into the sample containment chambers. The bucket includes a perimetric side wall having a predominant surface area that is contoured on opposite sides of the side wall to define exterior indentations in the predominant surface area; and a handle, including a pair of tongs shaped and disposed for insertion into the side wall indentations, and a telescopic handle joining the tongs. A radiation shield within the container shields the outside of the container from radiation derived in the activation region from the radiation source. |
abstract | A method for generating electromagnetic radiation, including the steps of: providing a series of adjacent electrode pairs arranged on a common dielectric substrate, the electrodes of each electrode pair substantially aligned on opposite sides of the common dielectric substrate; energizing a first electrode pair in the series of electrode pairs at an energizing time to produce a volume polarization distribution pattern within the common dielectric substrate; energizing a next adjacent electrode pair in the series of adjacent electrode pairs at a next energizing time to produce a variation of the volume polarization distribution pattern within the common dielectric substrate, the center of the next adjacent electrode pair located a distance from the center of the previous adjacent electrode pair, wherein the next energizing time is a time interval after the previous energizing time, the time interval less than the time for light to travel the distance between the centers of the previous and the next adjacent electrode pairs; and repeating the step of energizing the next adjacent electrode pair for subsequent electrode pairs in the series of adjacent electrode pairs to produce a continuous time-varying volume polarization distribution pattern within the common dielectric substrate. |
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description | In the drawings, like references indicate like parts. FIG. 1 schematically depicts a lithographic projection apparatus according to the invention. The apparatus comprises: a radiation system LA, Ex, IN, CO for supplying a projection beam PB of radiation; a first object table (mask table) MT for holding a mask MA (e.g. a reticle), and connected to first positioning means for accurately positioning the mask with respect to item PL; a second object table (substrate table) WT for holding a substrate W (e.g. a resist-coated silicon wafer), and connected to second positioning means for accurately positioning the substrate with respect to item PL; a projection system (xe2x80x9clensxe2x80x9d) PL for imaging an irradiated portion of the mask MA onto a target portion C (die) of the substrate W. As here depicted, the apparatus is of a transmissive type (i.e. has a transmissive mask). However, in general, it may also be of a reflective type, for example. The radiation system comprises a source LA that produces a beam of radiation. This beam is passed along various optical components comprised in an illumination system,xe2x80x94e.g. beam shaping optics Ex, an integrator IN and a condenser COxe2x80x94so that the resultant beam PB is of a desired cross-section and uniformly intense throughout its area. The beam PB subsequently intercepts the mask MA which is held on a mask table MT. Having passed through the mask MA, the beam PB passes through the lens PL, which focuses the beam PB onto a target portion C of the substrate W. With the aid of the interferometric displacement and measuring means IF, the substrate table WT can be moved accurately, e.g. so as to position different target portions C in the path of the beam PB. Similarly, the first positioning means can be used to accurately position the mask MA with respect to the path of the beam PB, e.g. after mechanical retrieval of the mask MA from a mask library, or during a scan. In general, movement of the object tables MT, WT will be realized with the aid of a long stroke module (course positioning) and a short stroke module (fine positioning), which are not explicitly depicted in FIG. 1. In the case of a waferstepper (as opposed to a step-and-scan apparatus) the mask table may be connected to a short stroke positioning module only, or it may just be fixed. The depicted apparatus can be used in two different modes: 1. In step mode, the mask table MT is kept essentially stationary, and an entire mask image is projected at once (i.e. a single xe2x80x9cflashxe2x80x9d) onto a target portion C. The substrate table WT is then shifted in the x and/or y directions so that a different target portion C can be irradiated by the beam PB; 2. In scan mode, essentially the same scenario applies, except that a given target portion C is not exposed in a single xe2x80x9cflashxe2x80x9d. Instead, the mask table MT is movable in a given reference direction (the so-called xe2x80x9cscan directionxe2x80x9d, e.g. the y direction) with a speed v, so that the projection beam PB is caused to scan over a mask image; concurrently, the substrate table WT is simultaneously moved in the same or opposite direction at a speed V=Mv, in which M is the magnification of the lens PL (typically, M=1/4 or 1/5). In this manner, a relatively large target portion C can be exposed, without having to compromise on resolution. In what follows, it will be assumed that the radiation system produces UV light with a wavelength of 157 nm. According to the invention, the spaces in the apparatus that are traversed by the illumination beam, both before and after it has passed through the mask, are flushed with a laminar flow of flushing gas. The flushing gas may be ultra-pure nitrogen (N2) or other gas or gas mixture sufficiently transparent to the illumination radiation used in the apparatus. N2 has an extinction coefficient, k, at 1 standard atmosphere of less than about 0.0001 per cm traversed, as compared to air at 1 standard atmosphere for which k is approximately 46 per cm traversed. The actual gas pressure in the beam path may be above atmospheric pressure, so that any leak results in an outflow of flushing gas rather than contamination by air, or below atmospheric to reduce beam absorption further. In critical areas, for example the beam delivery and illumination optics, the flushing nitrogen is provided at high purity, i.e. with an air contamination of less than 1 ppm. In less critical areas, such as the projection lens, a contamination of up to 10 ppm can be tolerated whilst in the reticle and wafer stages contamination levels of up to 100 ppm and up to 500 ppm respectively may be tolerable. FIG. 2 shows the mask stage of the lithographic apparatus according to the invention in greater detail than FIG. 1. It will be seen that the mask M is held in a recess in mask table MT, which can be manufactured from a ceramic material such as Zerodur (RTM) and is positioned by a drive system (not shown) during operation of the lithography apparatus. The mask table MT is closely sandwiched between the last element of the collimating optics CO, which generate the projection beam PB, and the first element of the projection lens system PL, which projects the projection beam PB, having traversed the mask M, onto the wafer W (shown in FIGS. 1 and 3). The mask stage is divided into zones or spaces 2 to 6 as follows: space 2 is between the final illuminator optics CO and mask table MT; space 3 is within the mask table MT above the mask M; space 4 is within the mask table MT, between the mask M and pellicle 13; space 5 is within the mask table MT below the pellicle 13; and space 6 is between the mask table MT and projection lens system PL. Each of the spaces is flushed with a substantially laminar flow of flushing gas provided from flushing gas supply 11 via respective flow regulators 112 to 116. At the other side of each space the flushing gas is removed to reservoir 12 via respective vacuum pumps 122 to 126. Reservoir 12 may be partitioned to allow controlled re-use of the gas in selected spaces and may include devices 12a to clean or scrub the recovered gas. To ensure laminar flow, the various spaces in the mask stage can be separated from one another. In particular, a thin sheet 14, e.g. of a material substantially transmissive to the employed radiation, such as CaF2 or fused SiO2, for example, is provided to cover the recess in the mask table MT and separate space 2 from space 3. Flow regulators 112, 113 and vacuum pumps 122, 123 are controlled to ensure that there is no, or only minimal, pressure differential between these two spaces to avoid loading sheet 14. Sheet 14 is arranged to be easily dismountable and replaceable during reticle exchange. Similarly, space 5, below pellicle 13, is closed off from space 6, between the mask table MT and projection lens PL, by a second thin sheet 15 which may be made of CaF2 or fused SiO2. Sheets 14 and 15 may also be made of MgF2, BaF2 or any other suitable material that transmits radiation at the wavelength used in the apparatus. Similar sheets may also be used to cover any irregularities or dead spaces in the system. For example, a third sheet 16 may be used to cover the non-flat surface of the first element of the projection lens system PL. Sheets 14 and 15 and their like form partitions to isolate parts of the beam path in which the laminar flushing gas flow is provided. It should be noted that sheets 14, 15 and 16 are provided to smooth the laminar flow of flushing gas and need not be gas-tight, nor necessarily form a gas-tight seal to the parts to which they are connected. To supply and remove the gas flow to spaces 3, 4 and 5, within the mask table MT, appropriate conduits are provided in the body of the mask table. When the mask table has been exposed to air, e.g. after a period of non-operation of the apparatus or after mask exchange, flushing gas is supplied for a short period before an exposure is taken to flush out any air that may have accumulated in non-flat parts of the mask table, e.g fiducials. In this embodiment a pellicle is provided and may be made of solid SiO2 or CaF2. Fused silica (SiO2) which has an improved transmission at 157 nm may also be used. Polymer pellicles are preferably avoided to avoid diffusion across them. In some embodiments of the invention the pellicle may be omitted altogether in which case the flushing gas supply is simplified. In any of the spaces, aerodynamic features such as small strips or fins may be provided as desired to smooth or guide the flushing gas flow and eliminate or control vortex production. The gas supply and evacuation conduits, particularly in spaces 2 and 6, are positioned to minimize the length of the gas flow to reduce the opportunity for mixing with air. FIG. 3 shows the wafer stage of the lithographic apparatus of FIG. 1. In the wafer stage there is only a single space to be flushedxe2x80x94between the last element of the projection lens system PL and the wafer W. To avoid having to provide a flushing gas path covering the entire range of movement of the wafer stage, the flushing gas supply outlets 17 and evacuation inlets 18 are mounted on the lower end of the projection lens system PL, either side of the final element. Outlets 17 and inlets 18 are respectively connected to the flushing gas supply 11 and reservoir 12 via flow regulator 117 and vacuum pump 127 respectively. The outlets 17 in particular, but also the inlets 18, may be provided with vanes to guide the flow of flushing gas. If not already flat, the final element of the projection lens system PL may be covered with a thin sheet as discussed above. The flow regulators 112 to 117 mentioned above may comprise static or controllable pressure or flow reducers and/or blowers as required to provide the necessary gas flow rates for the particular embodiment and the available gas supply. As described above, a laminar flow of flushing gas can be used to reduce absorption of the illumination beam in the moving parts of the lithography device without leading to excessive consumption of the flushing gas. Similar arrangements may also be used in the static components, such as the illumination beam generator and shaper and the projection lens system. However, it is simpler to seal static components than it is to seal moving components and it so may be more convenient to do so and maintain the static components under vacuum or with a static fill of transparent gas such as N2. While we have described above a specific embodiment of the invention it will be appreciated that the invention may be practiced otherwise than described. The description is not intended to limit the invention. In particular, it will be appreciated that the invention may be used in either or both the mask or substrate stage of a lithographic apparatus, and in any other type of apparatus employing a short wavelength radiation beam. |
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abstract | An arrangement and a method for irradiation of products by charged particles. The arrangement comprises a particle acceleration device and a radiation chamber, in which the products are irradiated from at least two sides, by that the particle beam is deflected with a scanning magnet and redeflected with redeflection magnets. The particle beam impinges on the products substantially parallel to the main normals of the surfaces of the products, which normals are positioned substantially perpendicular to the direction that the radiation axis has immediately before the deflection. An absorption means comprising a particle stopper is arranged in the space between the pole pieces of the redeflection magnets. A transport device for double-sided irradiation transports the products, which during transport are fixed to the transport device, through the irradiation arrangement. |
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claims | 1. A radiation phase-contrast imaging device comprising:an imaging system;the imaging system being composed ofa radiation source configured to irradiate radiation,a grating in which an absorber absorbing the radiation and extending in one direction is arranged in a direction perpendicular to the one direction, anda detection unit configured to detect a self-image of the grating generated by Talbot interference on a detection surface in which a detection element configured to detect the radiation is arranged in a matrix in a plane; anda position changing unit configured to change a relative position of the imaging system and an object such that a projection of the object moves linearly on the detection surface while keeping s positional relation of the radiation source, the grating, and the detection unit,wherein a longitudinal direction which is a direction along which the detection element on the detection surface of the detection unit is arranged is inclined with respect to an extending direction of the absorber of the grating, andwherein the imaging device further comprises a self-image generation unit configured to be operated when imaging is continuously executed while moving the imaging system and generate a plurality of self-images different in a reflecting position of a dark line of the grating based on detection data output from each portion of the detection surface. 2. The radiation phase-contrast imaging device according to claim 1,wherein a lateral direction which is a direction along which the detection element on the detection surface of the detection unit is arranged is inclined with respect to an arrangement direction of the absorbers of the grating. 3. The radiation phase-contrast imaging device according to claim 1,wherein the detection surface of the detection unit includes a rectangular region configured such that an array in which a stripe-shaped self-image of one cycle is reflected and the detection element is arranged in one row in the longitudinal direction is arranged in a lateral direction. 4. The radiation phase-contrast imaging device according to claim 1, further comprising a radiation source controller configured to make the radiation source irradiate the radiation every time a projection of the object moves by an amount corresponding to one detection element on the detection surface. 5. The radiation phase-contrast imaging device according to claim 1,wherein the grating includes a region in which an absorber absorbing the radiation and extending in the one direction is arranged in a direction perpendicular to the one direction, and a region in which an absorber absorbing the radiation and extending in an intersecting direction intersecting with the one direction are arranged in a direction perpendicular to the intersecting direction, andwherein both the regions are arranged in a direction along which the projection of the object moves on the detection surface. 6. A radiation phase-contrast imaging device comprising:an imaging system;the imaging system is composed ofa radiation source configured to irradiate radiation,a grating in which an absorber absorbing the radiation and extending in one direction is arranged in a direction perpendicular to the one direction,a detection unit configured to detect a self-image of the grating generated by Talbot interference on a detection surface on which a detection element configured to detect the radiation is arranged in a matrix in a plane; anda position changing unit configured to change a relative position of the imaging system and an object such that a projection of the object moves linearly on the detection surface while keeping a positional relation of the radiation source, the grating, and the detection unit,wherein a longitudinal direction which is a direction along which the detection element on the detection surface of the detection unit is arranged coincides with an extending direction of the absorber of the grating and is inclined with respect to a moving direction of the projection of the object on the detection surface, andwherein the imaging device further comprises a self-image generation unit configured to be operated when imaging is continuously executed while moving the imaging system and generate a self-image in which a center point of the detection element is arranged at equal intervals on a line segment perpendicular to a moving direction of the imaging system. 7. The radiation phase-contrast imaging device according to claim 6,wherein a lateral direction which is a direction along which the detection element on the detection surface of the detection unit is arranged does not perpendicularly intersect with the moving direction of the projection of the object on the detection surface. 8. The radiation phase-contrast imaging device according to claim 6,wherein on the detection surface of the detection unit, an oblique direction along which it advances by an amount corresponding to one detection element in a lateral direction as it advances from a given detection element in the longitudinal direction by an amount corresponding to three detection elements coincides with the moving direction of the projection of the object on the detection surface. 9. The radiation phase-contrast imaging device according to claim 8, further comprising a radiation source controller configured to make the radiation source execute irradiation of the radiation every time the projection of the object on the detection surface moves by 1/101/2 times a width of one detection element. 10. A radiation phase-contrast imaging device comprising:an imaging system;the imaging system is composed ofa radiation source configured to irradiate radiation,a grating in which an absorber absorbing the radiation and extending in one direction is arranged in a direction perpendicular to the one direction, anda detection unit configured to detect a self-image of the grating generated by Talbot interference on a detection surface for detecting the radiation; anda position changing unit configured to change a relative position of the imaging system and an object such that a projection of the object moves linearly on the detection surface while keeping a positional relation of the radiation source, the grating, and the detection unit,wherein an array configured by detection elements arranged in a inclined direction which is a direction inclined with respect to a longitudinal direction is two-dimensionally arranged by being arranged in a lateral direction perpendicular to the longitudinal direction on the detection surface of the detection unit, andwherein the inclined direction is inclined with respect to an extending direction of the absorber of the grating, andwherein the imaging device further comprises a self-image generation unit configured to be operated when imaging is continuously executed while moving the imaging system and generate a plurality of self-images different in a reflecting position of a dark line of the grating based on detection data output from each portion of the detection surface. |
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description | This application claims the benefit of priority under 35 U.S.C. § 119 (a) and (b) to U.S. Provisional Application No. 62/802,860, filed Feb. 8, 2019, the entire contents of which are incorporated herein by reference. A steam methane reformer (SMR) is an industrial apparatus that utilizes heat, pressure, and a catalyst to convert methane (CH4) and steam (H2O) into hydrogen (3H2) and carbon monoxide (CO). The CH4 and H2O react with each other in catalyst-filled tubes that are enclosed within a furnace which provides heat to the endothermic reaction that forms H2 and CO. In order for this endothermic reaction to proceed, a very high furnace temperature is required. Typically, the temperature will be within 700-1,000° C. (1,292-1,832° F.), while the pressure may typically range from 3 to 25 bar. Some companies manufacture thermocouples that can be inserted into SMR tubes, by which the temperature can be measured along the length of the tube. Knowing this inside-of-SMR-tube temperature is valuable, so that the SMR reaction can be monitored, the process can be optimized, and the tube failure can be predicted. Installing these inside-of-SMR-tube thermocouples can be challenging because they need to be installed before catalyst is installed around them. This invention outlines an apparatus by which the inside-of-reactor tube thermocouple is installed and centered, and then catalyst is loaded into the tube afterwards. A device for centering a temperature measurement device inside a tube reactor that will be filled with catalyst, including a single inflatable bladder mechanically and fluidically attached to a centering ring. 101=steam methane reformer tube 102=tube inner surface 103=centering ring 105=single inflatable bladder 106=temperature measurement device 107=pressurized gas conduit 111=single bladder positioning system (including centering ring 103 and single inflatable bladders 105) 112=catalyst 113=bottom (distal end) of the reactor tube 114=top (proximal end) of the reactor tube 116=centering ring tether 117=vibration device 118=compressed gas source 119=compressed gas source valve 120=catalyst sock 121=catalyst sock tether 122=compressed air bleed valve 123=distal end mesh disk Illustrative embodiments of the invention are described below. While the invention is susceptible to various modifications and alternative forms, specific embodiments thereof have been shown by way of example in the drawings and are herein described in detail. It should be understood, however, that the description herein of specific embodiments is not intended to limit the invention to the particular forms disclosed, but on the contrary, the intention is to cover all modifications, equivalents, and alternatives falling within the spirit and scope of the invention as defined by the appended claims. It will of course be appreciated that in the development of any such actual embodiment, numerous implementation-specific decisions must be made to achieve the developer's specific goals, such as compliance with system-related and business-related constraints, which will vary from one implementation to another. Moreover, it will be appreciated that such a development effort might be complex and time-consuming, but would nevertheless be a routine undertaking for those of ordinary skill in the art having the benefit of this disclosure. The term “about 1 foot” means as close to one foot as is reasonable given the conditions. As used herein, “about 1 foot” is defined as meaning 12 inches plus or minus 20%. Likewise, “about 5 feet” is defined as 60 inches plus or minus 20%. The term “essentially vertical orientation” means as close to vertical is reasonable given the conditions. The term “essentially vertical orientation” means sufficiently vertical such that the performance of the instant invention, from the perspective of one with ordinary skill in the art, is the same as if the tubes were precisely vertical. As used herein, ‘essentially vertical orientation” is defined as within 10 degrees from true vertical, preferably within 5 degrees of true vertical. As an overview, a method is provided that uses a sock catalyst loading technique. A centering ring is centered within the reformer tube using a single multiple inflatable bladder that will allow the catalyst to collect on the upper surface, then fall into the lower empty space of the reactor tube once deflated. To start, the bladder is positioned above the tube's catalyst layer. Or, if the tube is completely empty, the bladder is positioned above the end of the tube. The distance above the tube end (or catalyst layer) may vary, but about 12 inches is typical. The bladder is inflated, typically using nitrogen or dry service air. Then a “sock loading” technique is used to introduce the catalyst to the surface of the bladder. After the bladder is deflated and the catalyst fills the void beneath, the bladder is then repositioned, typically 1 foot above catalyst layer, and the process is repeated until the tube is fully loaded with catalyst. FIG. 1 represents a cross-sectional view of reactor tube 101 utilizing the instant device and method. The reactor tube may be a steam methane reformer (SMR) tube. Inflatable bladder 105 is attached to, and preferably spaced evenly around, the perimeter of centering ring 103. As indicated in FIG. 2 (cross-sectional view, inflated), for ease of explanation, this assembly will be referred to as single bladder positioning system 111. Within centering ring 103 are at least temperature measurement device 106 and pressurized gas conduit 107. Pressurized gas conduit 107 introduces and evacuates the gas used to inflate and deflate single inflatable bladder 105. Pressurized gas conduit 107 connects directly to single inflatable bladder 105, thus allowing it to be inflated and deflated. When inflated (as indicated in FIG. 5), single inflatable bladder 105 makes at least partial contact with tube wall inner surface 102 and acts to locate centering ring 103 centrally within tube wall inner surface 102. Temperature measurement device 106 is located inside of centering ring 103, and after the inflation of single inflatable bladder 105 is also positioned near the axial center of reactor tube 101. Although only one temperature measurement device 106 is indicated in the figures, it is understood that two or more temperature measurement devices 106 may be located inside centering ring 103 (not shown). This embodiment may be better understood with reference to FIGS. 1-10. Before the filling of reactor tube 101 with catalyst may commence, the tube is positioned in an essentially vertical orientation, and a sieve (not shown) has been located at distal end 113 of reactor tube 101, in order to keep catalyst 112 within reactor tube 101. It is preferred that the inside of reactor tube 101 be inspected to ensure that undesired objects or foreign material not be present. In some cases, a Boroscope (not shown) is used to determine that the inside of reactor tube 101 is clean. First, temperature measurement device 105 is attached to distal end mesh disk 123. This helps secure temperature measurement device in place when catalyst 112 is loaded above it. This keeps temperature measurement device 105 from migrating up reactor tube 101 during the subsequent steps. Starting at proximal end 114, temperature measurement device 106 is then inserted down the length of empty reactor tube 101. Temperature measurement device 106 and pressurized gas conduit 107 are inserted into the center of centering ring 103, and centering ring tether 116 is attached to centering ring 103. While two centering ring tethers 116 are shown in the figures, temperature measurement device 106 may be fitted with one or multiple centering ring tethers 116. Pressurized gas conduit 107 may be inserted through centering ring 103, or pressurized gas conduit 107 may not be inserted through centering ring 103. Pressurized gas conduit 107 may be left outside centering ring 102 if desired (not shown). Then, while holding and extending centering ring ether 116, lower single bladder positioning system 111 into reactor tube 101. Then lower single bladder positioning system 111 until it touches the bottom of reactor tube 101. At this time, it is recommended that centering ring tether 116 is marked to indicate the level at proximal end 114 of the tube, as a point of reference. Single bladder positioning system 111 is then raised a first distance H1 from the end of reactor tube 101. H1 is entirely discretionary, but a typical value would be about one foot. This distance may be determined from the mark previously made on centering ring tether 116. An additional reference mark may be made at this time on centering ring tether 116 with reference to proximal end 114, to indicate the new location of single bladder positioning system 111. Single inflatable bladder 105 is inflated, by closing valve 122 and opening valve 107, thereby at least partially contacting tube wall inner surface 102 and locating centering ring 103 centrally within reactor tube 101. Any available and suitable compressed gas source 118 may be used for inflating the bladders, such as compressed nitrogen or dry service or plant air. Temperature measurement device 106, which may be approximately the same length as reactor tube 101, is essentially centered within reactor tube 101. Catalyst sock 120 is loaded with a predetermined, or calculated amount of catalyst 112. Catalyst sock 120 is introduced into reactor tube 101 by conforming it around temperature measurement device 106 and pressurized gas conduit 107. It is recommended that catalyst sock 120 has approximately 2 inches of overlapping fold. As catalyst sock 120 is then lowered, tension must be kept on both centering ring tether 116 and pressurized gas conduit 107, in order to allow catalyst sock 120 to pass down reactor tube 101. Catalyst sock 120 is lowered until it reaches, or is in near proximity to, single bladder positioning system 111. It is important that the release point is close to single bladder positioning system 111 in order to reduce any damage to the falling catalyst. Catalyst sock 120 is then unloaded of catalyst 112. This is done by quickly tugging on the sock and releasing the overlapping fold. The empty sock is then removed from reactor tube 101. It is understood that this is an illustrative, but non-limiting, example of a method of sock loading a reactor tube. Other methods of sock loading known to the art may also be used with the instant method and device. As catalyst 112 is emptied, it collects on top of single inflatable bladder 105. Single inflatable bladder 105 is then deflated by closing valve 107 and opening valve 122. Catalyst 112 falls into the void below and is allowed to settle in and around centering ring 103 and temperature measurement device 106, thus maintaining their central position. In order to promote better catalyst packing, and to help avoid unwanted voids in the catalyst, reactor tube 101 may be vibrated 117. Catalyst sock 120 is then refilled with a predetermined amount of catalyst. Single bladder positioning system 111 is the raised a predetermined distance H2 and the process is repeated. Predetermined distance H2 may be any distance useful or meaningful to the installer, for example one foot. Again, this distance may be determined from the mark previously made on centering ring tether 116, and an additional reference mark may be made at this time on centering ring tether 116. This process is repeated to the desired location within reactor tube 101, which may be a desired distance from the top of the tube. Reactor tube 101 is now fully reloaded and full of fresh catalyst, and temperature measurement device 106 is centrally located and able to provide accurate and meaningful readings of temperature along the length of reactor tube 101. It will be understood that many additional changes in the details, materials, steps and arrangement of parts, which have been herein described in order to explain the nature of the invention, may be made by those skilled in the art within the principle and scope of the invention as expressed in the appended claims. Thus, the present invention is not intended to be limited to the specific embodiments in the examples given above. |
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042591538 | claims | 1. A device for removal of elements, namely, fuel assemblies and cans of a control and safety system, from a position within a core of a nuclear reactor, the elements having a first end forming a head and a second end opposite said first end, said device comprising: a hollow bar having a length longer than the length of an element to be removed, an open, core-facing end and an internal diameter sufficiently large that said hollow bar envelops said elements; means for moving said hollow bar into the core of the nuclear reactor; a main grip arranged in said hollow bar for axial movement relative to said bollow bar for gripping heads of elements to be removed; means for driving said main grip with respect to said hollow bar so that a gripped element is axially moved within said hollow bar; an auxiliary grip mounted on said core-facing end of said hollow bar for gripping a lowermost peripheral surface of said elements being removed; means for driving said auxiliary grip; said hollow bar being sufficiently long that the bar is movable into an extended position in which the bar substantially encompasses all of an element and has its core-facing end at the level of the second end of an element positioned within the core with said auxiliary grip reaching and gripping the peripheral surface at the second end of said element so that upward movement of said auxiliary grip frees a jammed element from the core without subjecting the element to tension forces, the other end of said bar being outside of the reactor. 2. A device as claimed in claim 1, which comprises a sleeve member coaxially fitted over said hollow bar and adapted for movement along said hollow bar; said auxiliary grip having a plurality of jaws with profiled external surfaces, said sleeve member enveloping said auxiliary grip and said jaws, said sleeve member having a core-facing end with a plurality of openings, the number of said openings corresponding to the number of said jaws, each of said openings being adapted to be positioned opposite one of said jaws, said sleeve member having profiled surfaces on opposite sides of said openings adapted to interact with said profiled surfaces of said jaws of said auxiliary grip to control the opening and closing of said jaws. 3. A device as claimed in claim 2, wherein said hollow bar and said sleeve member both have at least one slot extending longitudinally from the core-facing ends of the bar and the sleeve member, the length of said slots being not less than the length of said elements being removed, the location and width of said slots being determined by the shape and size of said elements to be removed and the arrangement of the elements in the reactor core. 4. A device as claimed in claim 2 further comprising means for axially moving said sleeve member with respect to said hollow bar so that said profiled surfaces of said sleeve member interact with said profiled surfaces of said jaws to open and close said jaws. 5. A device as claimed in claim 2 wherein the core-facing end of said sleeve member is profiled to interact with elements adjacent to an element to be removed to center the main grip with respect to the head of the element being removed. |
048044989 | summary | BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to a process for treating radioactive waste liquid and, more particularly, to a treatment of concentrated radioactive waste liquid containing a soluble salt as the main component. 2. Description of the Prior Art A solid of liquid low-level radioactive waste formed in a nuclear power plant or other facilities using radioactive rays is packed in a drum, where it is solidified with cement and is stored in this state in the nuclear power plant because delivery of the radioactive waste outside the nuclear power plant is not allowed. The volume reduction ratio of the radioactive waste is low in this cement solidification method and the number of cement-solidified drums stored in facilities increases year by year. Accordingly, from the viewpoint of economy of the storage space, various methods for increasing the volume reduction ratio of the radioactive waste have been developed. As the main liquid radioactive waste formed in the nuclear power plant, there can be mentioned a concentrated radioactive waste liquid. At the present, this concentrated radioactive waste liquid is dried and powdered to remove water occupying the majority of the volume, and solidification is effected with a hydraulic solidifying material, especially cement. However, the following problems arise when the concentrated radioactive waste liquid is dried, powdered and then solidified with the hydraulic solidifying material. In case of a concentrated radioactive waste liquid generated in a BWR power plant (wherein the main component is sodium sulfate (Na.sub.2 SO.sub.4)), calcium hydroxide (Ca(OH).sub.2) deposited when cement is set reacts with sodium sulfate (Na.sub.2 SO.sub.4) to form calcium sulfate (CaSO.sub.4), which in turn reacts with tricalcium aluminate (3CaO.Al.sub.2 O.sub.3) hydrate to form a swellable hydrate which deteriorates the solidified body. In case of a concentrated radioactive waste liquid generated in a PWR power plant (wherein the main component is sodium borate (Na.sub.2 B.sub.4 O.sub.7)), if the amount of the borate ion is increased relative to cement, the hydration of cement is hindered and a desirable solidified body cannot be obtained. This problem arises not only when cement is used but also when other hydraulic solidifying material is used. As means for solving the foregoing problems, there has been examined a method in which a concentrated radioactive waste liquid is subjected to an insolubilizing treatment (wherein sodium sulfate or sodium borate (Na.sub.2 SO.sub.4 or Na.sub.2 B.sub.4 O.sub.7), the soluble component in the concentrated radioactive waste liquid, is converted into an insoluble substance), and then, the concentrated radioactive waste liquid is dried, powdered and solidified. Particularly, insolubilization of the concentrated radioactive waste liquid generated in the PWR power plant is disclosed in Japanese Patent Laid-Open No. 186099/1983. However, from the results of experiments made by inventors of the present invention, it was found that powdering and solidification of the concentrated radioactive waste liquid are difficult if caustic soda (NaOH) formed in the insolubilizing treatment of the concentrated radioactive waste liquid is contained in the concentrated radioactive waste liquid to be subjected to powdering and solidification. SUMMARY OF THE INVENTION An object of the present invention is to provide a process for treating radioactive waste liquid wherein an insolubilized radioactive waste liquid containing a soluble salt as the main component can be stably solidified. Another object of the present invention is to provide a process for treating radioactive waste liquid wherein the radioactive waste liquid containing a soluble salt as the main component is subjected to an insolubilizing treatment, radioactive substances such as Cs and Sr are caught in a precipitate formed by the insolubilizing treatment, re-use of caustic soda (NaOH) formed in the insolubilizing treatment is facilitated and the radioactive substances are stably fixed in a solidified body formed by solidification. One characteristic feature of the present invention resides in that a soluble salt (sodium sulfate or sodium borate (Na.sub.2 SO.sub.4 or Na.sub.2 B.sub.4 O.sub.7)) contained as the main component in a concentrated radioactive waste liquid generated in the BWR power plant or the PWR power plant is insolubilized and precipitated, sodium hydroxide (NaOH) formed in the insolubilization is separated from the precipitate and the radioactive waste liquid slurry containing the precipitate is solidified with a hydraulic solidifying material. Another characteristic feature of the present invention resides in that in order to facilitate re-use of caustic soda (NaOH) formed in the insolubilizing treatment, an adsorbent capable of adsorbing the radioactive substances such as Cs and Sr is added to the radioactive waste liquid. Since the separated caustic soda (NaOH) is free of the radioactive substances, it can be easily utilized again, and since the radioactive substances are stably fixed in the solidified body, leakage of radioactivity from the solidified body can be greatly reduced. In the present invention, the solidifying material comprising an alkali silicate as the main component can be used. However, other hydraulic solidifying materials, for example, cement can be used. In the present invention, the BWR concentrated radioactive waste liquid (wherein the main component was Na.sub.2 SO.sub.4) can be treated and a soluble barium salt is used and added for the insolubilizing treatment. In the case where a PWR concentrated radioactive waste liquid (wherein the main component is Na.sub.2 B.sub.4 O.sub.7) is treated, a soluble calcium salt is used and added for the insolubilizing treatment. The intended effects are similarly attained also in this case. In the present invention, the adsorbent for adsorption for the radioactive substances is added to the radioactive waste liquid simultaneously with the addition of the additive for the insolubilization. Of course, the adsorbent for adsorption of the radioactive substances may be added before or after the addition of the additive for the insolubilization. The adsorbent for Sr is not limited to titanium chloride (TiCl.sub.2), but organic and inorganic titanium compounds and organic and inorganic zirconium compounds can be used. The adsorbent for Cs is not limited to copper ferrocyanide, but other metal ferrocyanides may be used. Furthermore, a zeolite may be used as the adsorbent instead of titanium chloride and copper ferrocyanide. |
abstract | An anti-scatter grid, a detector with such an anti-scatter grid and a radiation imaging system including such a detector with an anti-scatter grid are provided. The anti-scatter grid includes at least one grid wall. The parameters of the grid wall may be adjusted to arrive a uniform scatter-to-primary ratio. The parameters of the grid wall comprise thickness, height, shape, or position of the grid wall, or width of interspace between two grid walls. The detector includes the anti-scatter grid, at least one photosensor, and at least one scintillator. The radiation system includes a radiation generator, a radiation detector with the anti-scatter grid, and a processor. |
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039487237 | summary | BACKGROUND OF THE INVENTION 1. Field of the Invention This invention relates generally to liquid metal cooled nuclear reactors, particularly to equipment for handling nuclear fuel assemblies, and more particularly to an apparatus for isolating a particular reactor core position for easy insertion and withdrawal of fuel assemblies at that position. 2. Description of the Prior Art In any nuclear reaction which is intended to operate over an extended period of time, such as a reactor which is to be utilized for the generation of electrical power, the reactor facility should be constructed for removal and disposal of used or "spent" fuel elements and for recharging or refueling of the reactor with new or unused fuel elements. One particular problem which has been encountered with sodium-cooled fast reactors has been accurately locating the fuel subassembly to be removed. This problem comes about because of the relatively small size of the typical fuel assembly and the tight packing of these assemblies within the core region. It is further compounded by the fact that sodium is not transparent and thus the assemblies cannot be visually located. Another problem brought about by the tight packing of the fuel assemblies in the core is that of preventing the accidental withdrawal of adjacent assemblies when a desired assembly is being pulled from the core. Also, again due to the tight packing it is desirable to have some means for maintaining the core positions, from which an assembly has been withdrawn, open to facilitate the inserting of a new assembly in its place. SUMMARY OF THE INVENTION The present invention overcomes the prior art problems recited above by providing a device which is installed through the reactor closure head prior to refueling. The apparatus includes two separate sections, the handling plug/spreader assembly and the floor valve. The handling plug is a wedge shaped segment with an opening through it to permit passage of the refueling apparatus. During installation in the reactor the opening is closed by a shield plug, which is removed after the floor valve is installed. The handling plug with the spreader assembly attached replaces an identically shaped control rod island in a rotating plug in the reactor head. The handling plug when in place in the reactor serves as part of the reactor containment boundary to prevent contamination of the reactor room atmosphere during installation of the floor valve. The core spreader assembly consists of an elongated shroud attached to and extending downward from the handling plug and carrying at its lower end a radially floating, hexagonal spreader or camming tube. The attachment of the camming tube to the shroud is a self-adjusting universal type coupling that permits the tube to move radially to compensate for misalignment as the tube is inserted into the core. Mechanisms are provided to rotate the camming tube for angular alignment purposes and to translate it linearly for insertion into and withdrawal from the core. The upper end of the camming tube serves as a guide for leading the handling machine into alignment with the chosen subassembly. The floor valve sits on top of the handling plug and interconnects it with the fuel handling machine. It is basically a gate valve provided with shielding to protect operating personnel during handling of spent subassemblies. |
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claims | 1. A system comprising:a signal generator that generates a first signal having a first jitter characteristic of a first value;a signal processing circuit that has a first input, a second input, and an output, wherein the first input is coupled to receive the signal generated by the signal generator, wherein the signal processing circuit generates and supplies a second signal having a second jitter characteristic of a second value, and wherein the second value is less than the first value; anda field carrier having an input and an output, wherein the output of the signal processing circuit is coupled to the input of the field carrier, and wherein the output of the field carrier is coupled to the second input of the signal processing circuit. 2. The system of claim 1, wherein the field carrier is configured to carry a desired charge distribution, wherein the desired charge distribution has a first electromagnetic field, and wherein the system further comprises:a radiation source configured to generate a second electromagnetic field that interacts with the first electromagnetic field so as to produce a force on the charge carrier. 3. The system of claim 1, wherein the first value is at least one hundred times the second value. 4. The system of claim 1, wherein the field carrier is selected from the group consisting of: a set of wire cantilevers, and a set of traces on a printed circuit board. 5. A signal processing circuit comprising:a first coupling circuit having a first input, a second input, a first output, and a second output, wherein the first coupling circuit is adapted to receive a first signal onto the first input, and wherein the first signal has a first jitter characteristic of a first value; anda first delay circuit having an input and an output, wherein the input of the first delay circuit is coupled to the second output of the first coupling circuit, and wherein the output of the first delay circuit is coupled to the second input of the first coupling circuit. 6. The signal processing circuit of claim 5, wherein the first coupling circuit outputs a second signal having a second jitter characteristic of a second value, wherein the second value of the second jitter characteristic is less than the first value of the first jitter characteristic. 7. The signal processing circuit of claim 6, wherein the first value of the first jitter characteristic is at least ten times the second value of the second jitter characteristic. 8. The signal processing circuit of claim 5, further comprising:a second coupling circuit having a first input, a second input, and an output, wherein the first input of the second coupling circuit is coupled to the first output of the first coupling circuit; anda second delay circuit having an input and an output, wherein the output of the second delay circuit is coupled to the second input of the second coupling circuit, and wherein the output of the second coupling circuit is coupled to the input of the second delay circuit via a field carrier. 9. The signal processing circuit of claim 8, wherein the second coupling circuit outputs a third signal having a third jitter characteristic of a third value onto the output of the second coupling circuit, wherein the third value of third jitter characteristic is less than the second value of the second jitter characteristic. 10. The signal processing circuit of claim 9, wherein the first value of the first jitter characteristic is at least one hundred times the third value of the third jitter characteristic. 11. The signal processing circuit of claim 8, wherein the first coupling circuit is a first broadband coupler, wherein the second coupling circuit is a second broadband coupler, wherein the first delay circuit is a first delay line, and wherein the second delay circuit is a second delay line. 12. The signal processing circuit of claim 5, wherein the first signal has a wavelength, and wherein the first coupling circuit and the first delay circuit form a feedback loop having a length that is an integer multiple of the wavelength of the first signal. 13. The signal processing circuit of claim 12, wherein the feedback loop operates to reduce undesirable noise present in the first signal. 14. The signal processing circuit of claim 5, wherein the first signal is a periodic signal. 15. A method comprising:(a) receiving a first signal onto a first feedback loop circuit, wherein the first signal has a first characteristic;(b) supplying a second signal onto a second feedback loop circuit, wherein the second signal is generated and output by the first feedback loop circuit, wherein the second signal has a second characteristic, and wherein the first characteristic is at least ten times the second characteristic; and(c) supplying a third signal onto a field carrier, wherein the third signal is generated and output by the second feedback loop circuit, wherein the third signal has a third characteristic, and wherein the second characteristic is at least ten times the third characteristic. 16. The method of claim 15, wherein the first characteristic, the second characteristic, and the third characteristic are characterizations of signal noise. 17. The method of claim 15, wherein the field carrier is taken from the group consisting of: a set of wire cantilevers, and a set of traces on a printed circuit board. 18. The method of claim 15, wherein the first feedback loop circuit comprises a first coupling circuit and a first delay circuit, and wherein the second feedback loop circuit comprises a second coupling circuit and a second delay circuit. 19. The method of claim 15, wherein each of the first signal, the second signal, and the third signal is a periodic signal that has a wavelength. 20. The method of claim 15, wherein the first feedback loop circuit has a first feedback length that is an integer multiple of the wavelength of the first signal, and wherein the second feedback loop circuit has a second feedback length that is an integer multiple of the wavelength of the second signal. |
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description | 1. Field of the Invention The present invention relates to a plunger type rotatable and replaceable multi-source radiator, especially a multi-source radiator for replacement of radiation source in a rotatable movement. 2. Description of the Prior Art The multi-source radiator is suitable for equipment correlation and radiation experiment. Conventionally, there are two ways of selecting and transporting the inner radiation source of radiator, one of these is by gas passage and the other is by air or oil cylinder passage as shown in FIG. 1 and FIG. 2 respectively. The flaws of these conventional ways are as follows: (1) The radiation source shield 1 is used for storing and preventing radiations, but it may leak from gas pipe passage 14 (or cylinder passage 17). (2) The radiation source selection device 3 moves the selected radiation source 8 to the gas pipe passage 14 (or cylinder passage 17) and separates each radiation source to prevent mutual disturbance. However the radiation source room space 16 and the gas pipe passage 14 (or cylinder passage 17) may have shortage of shield due to limited space, especially, this phenomenon will take place easily for room contraction in case of numerous radiation sources. (3) When the air pump 15 moves the radiation source 8 to the radiation aperture 4, there would be a danger of the radiation source stuck in the gas pipe passage 14 (or cylinder passage 17), which is caused by inappropriate location of the radiation source selection device 3 and the orientation shift of the radiation source 8. (4) It is difficult to take out or reinstall the radiation source after it has been installed in the radiation source room 7. (5) The radiation source may not be capable of returning to the radiation source room for insufficient atmospheric pressure while in a power failure. Therefore, it has been long for engineers to solve above-mentioned flaws, and it is now to be solved by the present invention. The main object of the present invention is to provide a plunger type multi-source radiator with a radiation source rod designed to avoid radiation leakage caused by conventional gas pipe passage (or cylinder passage). Another object of the present invention is to provide a plunger type multi-source radiator with a radiation source rod design to avoid mutual disturbance between radiation sources for insufficient shield of the gas pipe passage (or cylinder passage). Another object of the present invention is to provide a radiation source rod replaceable with a positioning lever device and a mechanical arm to prevent the radiation source to be stuck or blocked. Another object of the present invention is to provide a radiation source rod replaceable from upper position to ease the replacement of radiation source. Another object of the present invention is to provide a method to return the radiation source back to the radiation source room automatically by gravitation, preventing the radiation source dwelled in radiation aperture area during power failure. In order to achieve the above-mentioned objects, the present invention provides a radiator consisting of one or plural radiation sources for use in equipment correlation and radiation experiment. The radiation source shield and the rotatable radiation source selection device form a main block of radiation sources, wherein a cavity formed on the radiation source shield for accommodating the rotatable radiation source selection device, a radiation aperture formed on left side of the main block for radiation, and bottom side of the main block kept hollowed for routing transmission lines. The main parts of the radiator may perform the following roles: a radiation source rod used to hold the radiation source between two shields to avoid radiating to both ends, reduce mutual obstruction and leakage; a rotatable radiation source selection device used to separate the radiation source and a positioning lever device used to move the radiation source exactly to the reserve radiation location; staircases shield type radiation aperture and the shields beneath the radiation source rod used may impede disturbances from other scattering radiation source to the selected radiation source at radiation aperture; the radiation source shield keep the radiation source in the radiator safely; a precise mechanical arm may move the radiation source to the radiation aperture or retract it to the reserve radiation location, even during power failure, it can be accomplished by gravitation. Therefore, the rotatable and replaceable plunger type multi-source radiator of the present invention is capable of conserving the radiation source in the radiator safely, positioning precisely, operating securely without leakage and mutual disturbance of radiation, and easy for assembly and disassembly. FIG. 1 and FIG. 2 are prior arts of multi-source radiator of the gas pipe passage type and the cylinder passage type, respectively. It is shown that possibility of mutual disturbance between radiation sources 8 is high for insufficient shield 1 of the conventional gas pipe passage 14 (or cylinder passage). During operation, a rotatable radiation source selection device 3 selecting the desired radiation source 8 and moving it to a position against a radiation aperture 4 for operation through air or cylinder pressure, respectively. As shown in FIG. 3, it illustrates an exploded view of the construction of the present invention. The multi-source radiator contains: a radiation source shield 1, plural shield cavities 2 for accommodation of plural plunger pipes 5 and radiation source rods 6, a rotatable radiation source selection device 3 for selecting a desired radiation source 8, a radiation aperture 4 for radiation source 8 to radiate, one or plural plunger pipes 5 for accommodation of plural radiation source rods 6, one or plural radiation source rod 6, one or plural radiation source box 7 for accommodation of radiation source 8, one or plural radiation source 8, a mechanical arm 9 for operating and moving radiation source rod 6, a positioning lever device 10 for precisely positioning the rotatable radiation source selection device 3, a control device 11 for controlling timing and positioning of the rotatable radiation source selection device 3 and the radiation source rod 6, a radiation position 12, and a reserve radiation location 13. In FIG. 4, it illustrates the radiation source rod 6 comprising: a handle 61 for operation with the mechanical arm 9, a packing gland 62, top shield 63 for preventing leakage of radiation, radiation source box 64, bottom shield 65 for preventing leakage of radiation, and a radiation source 66. The control device 11 may be used to set up exposure time with built-in timer (not shown), and a built-in displacement controller (not shown) may be used to control the rotatable radiation source selection device 3 to set up and select the radiation source 8, a CPU may be used to control the mechanical arm 9 in accordance with the built-in timer. When there is no operation, the radiation source rod 6 may fulfill the space inside the rotatable radiation source selection device 3 to avoid radiation leakage. During operation, firstly, the control device 11 selects the radiation source 8 and sets the exposure time, followed by pushing start button to make the rotatable radiation source selection device 3 move the chosen radiation source rod 6 to the proximity of the radiation position 12, and then the mechanical arm 9 lifts up the radiation source rod 6 to the adjacent position of radiation aperture 4 to start up radiation according to the built-in timer. The disturbance from other radiation sources can be obstructed with the top shield of standby radiation sources 8 and the bottom shield of the operating radiation source 8. The mechanical arm 9 may move the radiation source 8 to return to its reserve radiation location 13 to complete a radiation cycle when the radiation time is up, and it also can take the radiation source 8 back to the reserve radiation location 13 from the radiation position 12 by gravitation when an unexpected power failure occurred. Various additional modifications of the embodiments specifically illustrated and described herein will be apparent to those skilled in the art, particularly in light of the teachings of this invention. The invention should not be construed as limited to the specific form and examples as shown and described, but instead is set forth in the following claims. |
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abstract | An embodiment of the invention relates to an X-ray device, more particularly for phase-contrast imaging in the medical sector. In at least one embodiment, the X-ray device includes an X-ray radiation source, a coherence grid, a phase grid and an X-ray detector from a number of pixels arranged in a matrix, the pixels including a lens grid. |
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abstract | A nuclear fuel rod for fast reactors includes a metallic fuel slug coated with a protective coating layer. In embodiments, a nuclear fuel rod for fast reactors includes a uranium and zirconium fuel slug having a single protective coating which is an oxide layer having a thickness in the range of 0.5 μm to 100 μm, and the protective coating layer may be configured to (i) prevent interdiffusion between the fuel slug and a cladding tube during fast reactor operation, and (ii) prevent a cladding tube from thinning during fission operation in a fast reactor. |
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description | This application is a national phase of International Application No. PCT/EP2008/068173, entitled “COUNT RATE MEASUREMENT METHOD AND DEVICE”, which was filed on Dec. 22, 2008, and which claims priority of French Patent Application No. 07 60330, filed Dec. 24, 2007. Technical Field And Prior Art The present invention relates to a fission chamber count rate measurement device as well as a fission chamber calibration device that comprises the count rate device of the invention. The present invention also relates to a fission chamber count rate measurement method. Fission chambers are used to detect neutrons. A fission chamber contains the fissile material and a gas capable of ionising. Due to the effect of neutrons, the fissile material emits particles that ionise the gas. The quantity of gas ionised expresses the quantity of neutrons received in the fission chamber. Only a part of the fissile material, known as the effective mass, participates in the emission of the particles that ionise the gas. In practice, precise knowledge of the effective mass is necessary for the determination of the absolute physical magnitudes, which are the neutron flux or the indices of spectra. The calibration device of the invention makes it possible to measure the effective mass of the fissile isotope. At the moment, the calibration of fission chambers is carried out in nuclear reactors, either in thermal spectrum (or thermal column), or in fission spectrum. Numerous calibration methods have been developed in this context. Said methods all require the use and the availability of a research reactor. For reasons of safety, said methods require the implementation of experimental procedures that are laborious to implement and are, consequently, costly. Moreover, research reactors equipped with calibration devices are less and less numerous throughout the world, which explains the necessity of having to travel if it is wished to calibrate fission chambers. The calibration devices of the prior art have numerous drawbacks. The calibration device of the invention does not have these drawbacks. Description of the Invention Indeed, the invention relates to a count rate measurement device of at least one fission chamber that contains the fissile material, characterised in that it comprises: a measurement cell that contains the fission chamber, a neutron generator that emits neutrons in the form of periodic pulses towards the fission chamber, a measurement cable that collects, via a connector, a signal delivered by the fission chamber following the interaction of the neutrons with the fissile material, a neutron counter that delivers a neutron count signal emitted by the neutron generator, a processing system that delivers, over a predetermined time interval contained within a period of emission of the neutrons emitted by the neutron generator, a signal representative of the signal delivered by the fission chamber and a signal representative of the neutron count signal, and a circuit for computing the count rate of the fission chamber normalized with reference to the neutron count signal, from the signal representative of the signal delivered by the fission chamber and the signal representative of the neutron count signal. According to an additional characteristic of the invention: the measurement cell comprises a structure in which is formed a cylindrical cavity that opens via an opening into a wall of the structure, a first cylindrical jacket is placed on a wall of the cylindrical cavity, the first cylindrical jacket containing the fission chamber, the connector and a first part of the measurement cable, a second cylindrical jacket surrounds at a distance the first cylindrical jacket, the first and the second cylindrical jackets having, each, a first end fixed, in the structure, by a first ring situated on the side where the cavity opens into the wall of the structure and a second end fixed, in the structure, by a second ring, a third cylindrical jacket situated at the exterior of the structure is substantially aligned with the first cylindrical jacket, the third cylindrical jacket containing a second part of the measurement cable that extends the first part and a centering ring that maintains the cable in the second jacket. According to another additional characteristic of the invention, the device comprises: a sheet of material that covers the cylindrical jacket, and a hollow cylindrical structure placed between the sheet of material and the second cylindrical jacket. According to yet another additional characteristic of the invention, the sheet of material is a sheet of cadmium. According to yet another additional characteristic of the invention, the material that constitutes the hollow cylindrical structure placed on the sheet of material is made of boron and polyethylene According to yet another additional characteristic of the invention, the first cylindrical jacket and the second cylindrical jacket are separated by an air filled space. According to yet another additional characteristic of the invention, the structure in which the cavity is formed is made of graphite. According to yet another additional characteristic of the invention, the neutron generator is integrated in the structure of the measurement cell. The invention also relates to a calibration device intended to measure an effective mass of fissile material contained in at least one fission chamber, characterised in that it comprises a count rate measurement device according to the invention and a circuit for computing the effective mass of the fissile material from the count rate delivered by the count rate computing circuit. The invention also relates to a method of measuring the count rate of at least one fission chamber that contains the fissile material, characterised in that it comprises: an emission of neutrons in the form of periodic pulses towards the fission chamber so that the fission chamber delivers a signal that results from the interaction of the neutrons with the fissile material, a counting of the neutrons emitted to form a counting signal, a processing of the signal delivered by the fission chamber and the counting signal over a predetermined time interval contained within a period of emission of the neutrons emitted to deliver a signal representative of the signal delivered by the fission chamber and a signal representative of the counting signal, and a computation of the count rate of the fission chamber normalized with reference to the counting signal from the signal representative of the signal delivered by the fission chamber and the signal representative of the counting signal. The count rate measurement device of the invention comprises a neutron generator operating in pulsed mode. The device of the invention advantageously makes it possible: to obtain the calibration values with a precision equivalent to that obtained in reactor, and to obtain calibrations in various neutron spectra. Advantageously, the materials and dimensions of the different components that constitute the measurement cell in which is placed the fission chamber may be determined by a design method compatible, at the choice of the designer, with a fast spectrum or with a thermal spectrum of the interrogating neutrons of the fission chamber. A fast neutron spectrum is defined in that 99.9% of the neutrons have an energy above 1 MeV. A thermal neutron spectrum is defined in that 99.9% of the neutrons have an energy below 0.625 eV. For each type of neutron spectrum, the devices are adapted, for example, to fission chambers of 1.5 mm, 4 mm or 8 mm diameter. FIG. 1 represents a measurement cell that contributes to the count rate measurement device of the invention. The measurement cell 1 is constituted of an enclosure 2 that contains a material 3 in which is formed a cavity 4 and a neutron counter K. The material that constitutes the enclosure 2 is, for example, made of polyethylene and the material 3 is, for example, made of graphite. A longitudinal cavity 5 capable of accommodating a fission chamber is formed inside the material 3. A neutron generator 6 is placed, in the material 3, near to the cavity 5. The cylindrical cavity 5 opens into the cavity 4 by an opening O. In the embodiment represented in FIG. 1, the neutron counter K is placed next to the enclosure 2. The invention also relates to the case where the neutron counter is situated in the enclosure 2. FIG. 2 represents a partial sectional view of a first example of structure that contributes to the count rate measurement device of the invention. The structure represented in FIG. 2 is intended to obtain a fast neutron spectrum. The cavity 5 contains two cylindrical coaxial jackets 8, 9, the jacket 8 surrounding the jacket 9. The jackets 8 and 9 are, for example, made of stainless steel and have a thickness of 1 mm. A sheet of material 13, for example a sheet of cadmium of lmm thickness, covers the exterior face of the cylinder 9. The purpose of the sheet of material 13 is to capture the thermal neutrons, in other words the neutrons whose energy is below 0.625 eV. A block of material 10 is placed in the space that separates the sheet of material 13 and the jacket 8. The material 10, for example made of boron and poylethylene, has a thickness equal, for example, to 16 mm. Two centering rings 15 and 16 maintain and align the cylindrical jackets 8 and 9 in the cavity 5. A stop B closes the cavity on the side of the centering ring 16. The fission chamber CH is placed in the cylindrical jacket 9. A first end of the fission chamber is connected to a connection component 12 that collects the electrons created due to the ionisation of the gas contained in the chamber. This first end of the fission chamber is placed at a distance D from the opening O, the other end of the chamber being placed at a distance d from the stop B. The connection component 12 is connected to a rigid coaxial cable 11. A cylindrical jacket 17, for example a jacket made of stainless steel of 1 mm thickness, aligned with the cylindrical jacket 9, is placed in the cavity 4. A centering ring 14 maintains the rigid coaxial cable 11 in the cylindrical jacket 17. A connector connects the rigid coaxial cable 11 to a flexible measurement cable 7 that transmits the signal to processing circuits (not represented in FIG. 2; see FIG. 3). The guiding and positioning system constituted by the components 14, 15 and 16 advantageously guarantees a good reproducibility of the axial position of the chamber CH. The precision obtained for this position may be, for example, of the order of 1 mm, or even less. The materials and dimensions of the structure of the invention represented in FIG. 2 are preferentially obtained by means of the Monte-Carlo MCNP (MCNP for “Monte-Carlo N-Particle”) neutron computation code. It is precisely by this computation code that the above mentioned materials and dimensions have been obtained. However, other materials having equivalent characteristics could also be chosen for the constitution of this structure. The choice of these other materials would then imply a different dimensioning to obtain substantially equivalent performances. The above mentioned materials make it possible however to produce a calibration device having “acceptable” dimensions, in other words a device neither too voluminous nor too bulky. The choice of stainless steel for the cylindrical jackets 8, 9 and 17 makes it possible to ensure an excellent rigidity to the whole of the device and to guarantee its resistance to wear. The choice of boron and poylethylene is justified by the good resistance of this material to ageing, by its efficiency in terms of capture of thermal neutrons and by its low cost. The centering device 14, 15, 16 of the fission chamber is specific for each diameter of chamber studied. The centering rings 14, 15, 16 and the stop B are, for example, made of stainless steel. The diameters of the centering rings and the machining of the stop B are adapted to the diameter of the rigid coaxial cable 11. The centering device makes it possible not only to control the axial position and the radial centering of the chamber in the device but, also, allows the fission chamber to slide longitudinally in order to optimise the measurement position (and search for the position corresponding to the maximal count rate). In fact, it is the assembly constituted by the fission chamber CH, the connector 12 and the coaxial rigid 11 that slides in the jacket 9. The centering of this assembly is ensured by the ring 14. In the structure described above, only the neutrons not having undergone slowing down/thermalization in the graphite of the cell and in the boron and poylethylene penetrate inside the fission chamber. Thus, the fission chamber only sees the fast neutrons emitted by the generator 6, in other words the neutrons that have not undergone interaction. FIG. 3 represents a partial sectional view of a second example of structure that contributes to the count rate measurement device of the invention. The structure of FIG. 3 is adapted to obtaining a thermal neutron spectrum. The cavity 5 comprises all the constituent elements already described with reference to FIG. 2, with the exception of the block of material 10 and the sheet of cadmium 13. The space between the jackets 8 and 9 is here filled with air. In the same way as previously, the position of the fission chamber may be adjusted longitudinally by means of, for example, sliding means mentioned previously. The neutrons that have come from the generator 6 can here penetrate into the fission chamber whatever their energies. However, these neutrons pass through beforehand a thickness of graphite between, for example, 0 cm and around 40 cm according to the position occupied by the fission chamber in the jacket 9, which makes it possible to discriminate their energy as a function of their instant of arrival at the level of the fission chamber, in other words as a function of the thickness of graphite traversed. By way of non limiting example, the computations worked out from the Monte-Carlo MCNP4C2 code have shown that more than 99.9% of the neutrons emitted by a neutron generator, the firing frequency of which is 125 Hz are, after each firing, thermal neutrons in a time range between 700 μs and 3500 μs, whatever the axial position of the chamber in the calibration device. FIG. 4 illustrates a schematic diagram of the fission chamber count rate measurement device of the invention. The measurement device comprises: a measurement cell 1 as described above and in which are integrated a fission chamber CH, neutron generator 6 and a neutron counter K, a processing system ST of the signals delivered by the fission chamber CH and by the counter K and that delivers, on the one hand, a signal representative of the signal delivered by the fission chamber and, on the other hand, a signal representative of the signal delivered by the counter K, and a computing circuit 34 that calculates the count rate C of the fission chamber normalized in relation to the signal delivered by the counter K from the signals delivered by the processing system ST. The processing system ST comprises: a preamplifier 18 that amplifies the signal delivered, via the measurement cable 7, by the fission chamber CH, an amplifier 20 connected to the preamplifier 18 by a multi-conductor cable 19 that transmits towards the fission chamber a high voltage HT and a low voltage BT, an electronic circuit 21 connected by a cable 27 to the neutron generator 6, an acquisition circuit 32, which comprises an amplifier 22, an acquisition card 23 and a high voltage circuit 24, the amplifier 22 receiving, via a cable 26, the signal delivered by the counter K and, via an electrical connection 33, the high voltage delivered by the circuit 24, the cable 26 supplying in high voltage HT0 the counter K, the acquisition card 23 receiving, via an electrical connection 29, the signal delivered by the amplifier 20 and, via an electrical connection 28, the signal delivered by the electronic circuit 21, the amplifier 22 delivering the signal representative of the signal delivered by the counter K and the acquisition card 23 delivering the signal representative of the signal delivered by the fission chamber. By way of non limiting example, FIG. 4 represents a device containing a single fission chamber. However, the invention more generally relates to the case where the device contains N fission chambers, N being an integral number greater than or equal to 1. FIG. 5 represents a schematic diagram of fission chamber calibration device of the invention. In addition to the components mentioned with reference to FIG. 4, the calibration device represented in FIG. 5 comprises a computing circuit 35, which calculates the effective mass of the fissile material contained in the fission chamber from the normalized count rate C delivered by the computing circuit 34. An example of computation method implemented by the circuit 34 is given in the remainder of the description. Other computation methods could however be envisaged. The computation method will be described in a general case where N fissile masses contained in N different fission chambers are calculated simultaneously. Each fission chamber contains a principal isotope and impurities. The matrix column [m] formed from the effective masses of fissile material contained in the N fission chambers is written:[m]=[C]·I([a]×[σmφ]) (1), where [C] is the matrix of the normalized count rates (or number of hits per second) of N fission chambers, [a] is the matrix of the isotopic analyses of N deposits of fissile material, normalized in relation to the principal isotopes, [σmφ] is the matrix column of the integral mass cross sections of macroscopic fission normalized in relation to counter K (“mass cross section” is herein taken to mean a cross section relative to a mass unit and not to a nucleus), and where the symbol “·I” represents the matrix division operator and the symbol “×” represents the matrix multiplication operator. The matrix [C] is constructed from measurements delivered by the circuit 34. The coefficients of the matrix [a] are determined for each fissile material of each fission chamber, in a manner known per se, by isotopic analysis. For a fission chamber of given exterior dimensions (diameter, length), the matrix [σmφ)] is invariant for a given position of the fission chamber in the measurement cell and for a given calibration configuration of the calibration device (fast or thermal neutrons and time range linked to the nature of the spectrum). It is then possible to determine the matrix [σmφ] by means of a particular matrix [σmφ]0 that corresponds, for the above mentioned invariance conditions, to deposits of fissile material of known effective mass. It then follows:[σmφ]0=[a]0−1×[m]0−1×[c]0 (2)where the coefficients of the matrices [a]0−1 and [m]0−1 are known and the coefficients of the matrix [C]0 are the count rate measured for the deposits of fissile material of known effective mass. The matrix [m] is then written:[m]=[C]·I([a]×[σmφ]0) (3) In addition to the matrix [m], the computing circuit 35 also calculates the matrix of the variances var(m), where var(m) represents the variance of the effective mass m. The equation of the matrix of the variances is explained below. The above mentioned equation (2) may be written in a generic manner in the following way:[xij]=[Aij]×[Bij]×[Cij],where i is the index relative to the rank of the lines of the matrix and j the index relative to the rank of the columns of the matrix. By making the hypothesis of the independence of terms Aij, Bij and Cij, in other words assuming the independence of uncertainties on [C], [σmφ] and [a], the derivation of the uncertainty leads to:[var(Xij)]=[var(Aij)]×[(Bij)2]×[(Cij)2]+[(Aij)2]×[var(Bij)]×[(Cij)2]+[(Aij)2]×[(Bij)2]×[var(Cij)] The variances of the matrices [σmφ]0 and [m] are then written, respectively, as follows:var[σmφ]0=[[[m]0×[a]0]ij2]0−1×{var[c]0+([mij2]0×var[a]0+var[m]0×[aij2]0)×[σmφij2]0}andvar[m]={var[C]+[mij2]×(var[a]×[σmφij2]0+[aij2]×var[σmφ]0)}·I{[([a]×[σmφ])ij2]} In the above formulas, the notation [Zij2] represents the matrix constituted of the terms zij to the power 2, zij being the coefficient of the line of rank i and of the column of rank j of the matrix [Z]. Generally speaking, a fission chamber of principal isotope i contains impurities. In practice, U-234 and U-236 impurities are often present in negligible quantity in uranium U-233, U-235 or U-238 chambers and thus do not pose any problem. In the case Pu-238 chambers for example, the impurity U-234 is a product of the radioactive decay of Pu-238 with a period of 87.7 years. If a sufficiently recent Pu-238 chamber is employed, the quantity of U-234 will thus be negligible. In the cases where it is not possible to disregard the impurities, the computation method advantageously takes into account their influence. The calculated coefficients of the matrix [m] are then effective equivalent masses that take into account, apart from the effective masses of the principal isotopes, the effective masses of the impurities present in the fission chamber. By way of non limiting example, the expression of an effective equivalent mass of principal isotope Pu-238 that contains U-234 impurities will now be given. The equivalent number Neq of nuclei of Pu-238 isotope contained in a fission chamber is calculated with the following equation: N eq = ( N 4 × σ 4 , c σ 8 , c + N 8 ) where: N4 is the number of nuclei of U-234 contained in the chamber and known by analysis, N8 is the number of nuclei of Pu-238 contained in the chamber and known by analysis, σ4,c is the microscopic fission cross section of the impurity U-234, computed, for example, with the MCNP4C2 code, in the measurement conditions (counting time interval and fast neutron spectrum or thermal studied), σ8,c=microscopic fission cross section of Pu-238, computed, for example, with the MCNP4C2 code, in the measurement conditions (counting time interval and fast or thermal nature of the neutron spectrum studied). The effective equivalent mass meq of Pu-238 that is taken into account as coefficient of the matrix [m] is then given by the following formula: m eq = m 4 × 238 234 × σ m 4 c σ m 8 c + m 8 where m4 is the effective mass of U-234 in the chamber, 238 is the mass number of Pu-238, 234 is the mass number of U-234, σm4c is the microscopic mass fission cross section of U-234, computed, for example, with the MCNP4C2 code, in the measurement conditions (counting time interval and fast or thermal nature of the neutron spectrum studied), σm8c is the microscopic mass fission cross section of Pu-238, computed, for example, with the MCNP4C2 code, in the measurement conditions (counting time interval and fast or thermal nature of the neutron spectrum studied), m8 is the effective mass of Pu-238 in the chamber. The measurement of a fission chamber count rate by means of a device such as that represented in FIG. 4 will now be described. The determination of the count rate comprises the following principal steps: establishment of measurement conditions specific to the fast or thermal nature of the neutron spectrum; implementation of a measurement methodology. The establishment of the measurement conditions comprises: a determination of the optimal longitudinal position of the fission chamber, in relation to a count rate of the chamber, the optimal position retained corresponding to a maximum count rate, and thus to a minimum counting time for a given statistical counting uncertainty, a regulation of the acquisition time range and the number of acquisition passes adapted to the performances and specificities of the neutron generator, a regulation of the time of acquisitions relative to a statistical uncertainty objective of 1% on the count rate. The measurement methodology implemented comprises the following steps: a regulation in PHA mode (PHA for “Pulse Height Analysis”) of a discrimination threshold of the background noise contained in the acquisition signal of the fission chamber, an association of the discrimination threshold with the sought after calibration magnitude (effective mass), a transposition of regulation in PHA mode to acquisition in MCS mode (MCS for “Multi-Channel Scaler”) of the dynamic signal of the fission chambers during the operation of the neutron generator, and a monitoring of the measures obtained so as to dispense with operating fluctuations of the neutron generator from one measurement to the next. The different points mentioned above will now be made clear. Determination of the Optimal Position of the Chamber Preferably, it is advisable firstly to determine the optimal position of the fission chamber in the measurement cell so as to reduce as best as possible the acquisition time needed to obtain a statistical uncertainty of 1% on the count rates obtained. This step is not obligatory but it is recommended to save time and reduce as much as possible the wear of the neutron generator. FIG. 6 represents, in the case of a fast neutron spectrum, an example of curve of the normalized fission rate C measured as a function of the position that a fission chamber occupies in the measurement cell (distance D of the chamber in relation to the opening O). In the example chosen, the fission chamber contains around 100 μg of uranium-235 and the distance D that separates the chamber from the opening O varies from 5cm to 40cm. By way of indication, the acquisition time of each measurement is 800s, which makes it possible to obtain countings ranging between 2000 hits and 3500 hits over a region of interest between 30 μs as and 230 μs. Indeed, due to the cutting off of the thermal neutrons in the boron and poylethylene, no neutron reaches the fission chamber outside of this time range. The position retained is that corresponding to the maximum count rate. In the example of FIG. 6, the optimal position obtained is situated 23 cm from the opening O of the cavity 5. This position is then retained for all of the calibration measurements in fast neutron configuration. FIG. 7 represents an example of curve of the normalised fission rate C measured as a function of the position that the fission chamber occupies in the measurement cell in the case of a thermal neutron spectrum. In the example chosen, the fission chamber contains around 100 μg of uranium-235 and the position of the distance D varies from 0 to 40 cm. By way of indication, the acquisition time of each of the measurements is 560 s. This time makes it possible to obtain countings ranging between 10000 hits and 30000 his over a region of interest ranging between 700 μs and 3500 μs. This time range corresponds to the instants during which the population of thermal neutrons of energy below 0.625 eV is greater than 99.9%. The maximum count rate is here observed at the position of 5 cm. This position is then retained for all of the calibration measurements in thermal configuration. In all of the devices, the precision associated with the positioning of the fissile deposit of the fission chamber is of the order of 3 mm, including the uncertainty on the positioning of the fission chamber in the device and the uncertainty on the positioning of the deposit in the fission chamber. These imprecisions lead to an uncertainty of the order of 0.1% on the count rate of the fission chamber, which is virtually negligible. Acquisition Range-Number of Passes-Acquisition Time The operating and acquisition parameters of a measurement are the following: the firing frequency of the neutron generator, for example a frequency of 125 Hz, which corresponds to a time between two pulses of the neutron generator equal to 8000 μs; the number of passes, in other words the number of firings of the neutron generator, which it is necessary to adapt to the minimal counting integral sought, for example 10000 hits, over a given time range, namely, for the above mentioned frequency of 125 Hz, the range 30 μs-230 μs in fast neutron devices and the range 700 μs-3500 μs in thermal neutron devices, the acquisition time, which is equal to the number of passes multiplied by the time between two pulses of the neutron generator (the acquisition time varies as a function of the nature and the deposits contained in the fission chamber and is typically equal to ten minutes for a thermal neutron device and to one hour for a fast neutron device when the fission chamber contains of the order of 100 μg of actinide). The measurement methodology will now be described. Regulation in PHA Mode of the Discrimination Threshold of the Background Noise—Association of this Discrimination Threshold with the Sought after Calibration Magnitude According to an improvement of the invention, it is provided not to take into account, in the signal delivered by the fission chamber, the electronic background noise or the noise due to particles or radiation other than neutrons. As illustration, FIG. 8 represents an example of spectrum S of fission chamber in which noise lines appear. Said noise lines are present in the first channels of the signal. The strategy adopted to eliminate these noises is to set a discrimination threshold making it possible to exclude the background noise contained in the first channels of the signal and, consequently, to only integrate the useful part of the signal corresponding to the neutron reactions. The method employed consists in carrying out an acquisition of the signal of the fission chamber in PHA mode (linking the number of hits to the amplitude of the signal), then to link the discrimination threshold to the shape of the signal that is specific to the fission chamber, since this shape is relative to the energy deposited by the fission products in the chamber. A counting integral is associated with the discrimination threshold. A value of the calibration magnitude (i.e. a value of effective mass of the fissile deposit in the fission chamber) is then associated with this counting integral. By way of indication, the method comprises the following steps: a pinpointing of the channel Cmax corresponding to the maximum counting of the spectrum, a computation of the average of the countings over, for example, ten channels situated on either side of the channel Cmax, i.e. Vav this average value, a pinpointing, on the spectrum, of the channel corresponding to the value Vav/2, a linear regression around the zone situated between five points that precede and five points that follow Vav/2, a determination of the straight line equation obtained due to the linear regression, by means of the straight line equation thereby obtained, a determination of the channel R corresponding to Vav/2, and the adoption of a precision, for example, to two decimal places (it appears that the channel R is uniquely a function of the spectrum delivered and does not vary as a function of the nature of the neutron spectrum), a computation of the value of the channels corresponding to a fraction of the channel R (for example 0.4 R, 0.5 R and 0.6 R) while always keeping the same number of decimal places for the precision, an identification of the discrimination threshold as being the fraction of channel R corresponding as best as possible to the “valley” channel Cv (for example 0.5 R (cf. FIG. 8)), a computation of the counting integral between the discrimination threshold and the end of counting channel Cf (see FIG. 8). The channel Cf is characterised in that all of the channels greater than Cf correspond to a zero count rate. There is obviously interest to favour a maximum of useful signal, in other words to retain an integration threshold as close as possible to the “valley” channel Cv (see FIG. 8). The method of determination of the discrimination threshold is not unique, since other methods may be used. However, this method makes it possible a priori to dispense with any change to the measurement conditions (gain of the amplifier, shaping constant of the signal, energy of the incident neutrons, etc.). Through experience, it adds on average an uncertainty of 0.5% on the measured count rates. Transposition of the Regulation in PHA Mode to the Acquisition in MCS Mode of the Dynamic Signal As seen previously, obtaining calibration magnitudes necessitates only considering the signal from the fission chambers over a certain time range lying between two firings of the neutron generator, in order to meet the objectives in terms of purity of the neutron spectrum. The acquisition of the signals from the fission chambers in MCS mode (counting scale ranking the number of events as a function of time) is thus necessary, so as to be able to follow the signal over time. It is thus advisable to regulate the lower threshold of the discriminator so that it corresponds perfectly with the chosen integration threshold (see previous paragraph). To do this, it suffices to multiply the integration threshold defined beforehand in PHA mode for each chamber by the magnitude G such that:G=Range in volts of the discriminator/Number of acquisition channelsto obtain in volts the regulation value of the lower threshold of the discriminator. By way of non-limiting example, and in a manner known per se, the number of channels for the analysis of the measurements is equal to 1024 and the range in Volts of the discriminator is equal to 10. Since the determination of the lower discrimination threshold is exact, it may then be considered that the associated uncertainty is zero. Monitoring of Measurements So as to dispense with operating fluctuations of the neutron generator from one measurement to another (flux of emerging neutrons or wear of the generator), it is necessary to monitor the neutron emission of the neutron generator, so as to normalize all of the measurements made in reference to an identical operation of the neutron generator. It is for this purpose that the calibration device comprises a neutron counter K. The counter K may be positioned in the interior or the exterior of the measurement cell. By way of non limiting example, the counter K is represented outside of the measurement cell in FIG. 1. The monitoring may be carried out by means of any type of neutron detector. By way of non limiting example, the counter K is a helium counter 3 positioned near to, for example several tens of centimetres, the measurement cell. Throughout the count rate measurement procedure and, consequently, the calibration procedure, it is preferable not to displace the counter K in relation to the fission chamber and to the neutron generator. The principle of the monitoring is to record systematically and in a synchronised manner the counting of the counter K and the counting of the fission chamber. All of the measurements delivered by the fission chamber are then normalized with reference to the average counting of the counter K (circuit 34). It may then be considered, by making the hypothesis that the signal delivered by the counter K is stable over time, that the only uncertainties to take into account are the statistical uncertainties of counting of the chamber. Validation of the Calibration Device of the Invention A validation of the count rate measurement device of the invention will now be described on the basis of measurement results illustrated by FIGS. 9-12. An example of count rate CCH as a function of time delivered by a fast neutron device uranium-235 fission chamber is given in FIG. 9. It may be noted that the signal follows overall the emission pulse of the neutron generator in the time interval 30 μs-230 μs. The signal is indeed cancelled beyond 230 μs, which validates the quality of the fast device as regards stopping the neutrons thermalized in the surrounding graphite. For thermal neutron devices, two examples of count rate CCH measurements obtained with neptunium-237 and uranium-235 fission chambers are given respectively in FIGS. 10 and 11. It may be noted, for the neptunium-237 fission chamber, that the signal follows initially the emission pulse of the neutron generator, lying in the time interval 30 μs-230 μs, then cancels out, which signifies that beyond the emission of the pulse of the neutron generator, no fast neutron is any longer observed in the thermal devices (in particular over the time range 700 μs-3500 μs retained in the study), due to the fact that neptunium-237 has a zero fission cross section in the thermal domain. It may further be noted that the signal from the uranium-235 fission chamber (fissile isotope to the thermal neutrons) initially follows the emission pulse of the neutron generator, lying in the time interval 30 nμs-230 μs, then continues to evolve without cancelling itself out as the thermal neutrons arrive at the level of the detector after having passed through the graphite of the measurement cell. The above elements validate the conception of the calibration cell by means of Monte Carlo MCNP4C2 code computations, as has been mentioned above, and for which it has been agreed to analyse all of the measurements over the time interval 700 μs-500 μs, corresponding to the interrogation of the fission chamber by thermal neutrons of energy below 0.625 eV in more than 99.9% of cases. By way of illustration, FIG. 12 represents the count rate CK detected by a helium 3 counter used as monitor for all of the measurements. The counter, positioned at the rear of the cell, firstly sees arrive the fast neutrons emitting during the pulse of the generator then, in a delayed manner, the neutrons more or less thermalized in the graphite and the polyethylene of the measurement cell. By way of indication, counting integrals greater than or equal to 10000 hits (relative to a statistical uncertainty of 1%) have been obtained over around one hour in fast neutron device and over around ten minutes in thermal neutron device, from a fission chamber containing of the order of 100 μg of fissile material and a neutron generator operating in pulse mode at a frequency of 125 Hz and with a neutron emission of 3.109n.s−1. These elements also validate the conception of the devices. |
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description | The present application relates to nuclear fission reactors, and systems, applications, and apparatuses related thereto. Illustrative embodiments provide for the operation and simulation of the operation of fission reactors, and including the movement of materials within reactors. Illustrative embodiments and aspects include, without limitation, nuclear fission reactors and reactor modules, including modular nuclear fission reactors and reactor modules, nuclear fission deflagration wave reactors and reactor modules, modular nuclear fission deflagration wave reactors and modules, methods of operating nuclear reactors and modules including the aforementioned, methods of simulating operating nuclear reactors and modules including the aforementioned, and the like. The foregoing summary is illustrative only and is not intended to be in any way limiting. In addition to the illustrative aspects, embodiments, and features described above, further aspects, embodiments, and features will become apparent by reference to the drawings and the following detailed description. The present invention will now be described with reference to the accompanying drawings. In the drawings, like reference numbers may indicate identical or similar elements. Additionally, the left-most digit(s) of a reference number may identify the drawing in which the reference number first appears. Introduction In the following detailed description, reference is made to the accompanying drawings, which form a part hereof. In the drawings, generally similar symbols identify similar components, unless context dictates otherwise. The illustrative embodiments described in the detailed description, drawings, and claims are not meant to be limiting. Other embodiments may be utilized, and other changes may be made, without departing from the spirit or scope of the subject matter presented here. It is to be appreciated that the Detailed Description section, and not the Summary and Abstract sections, is intended to be used to interpret the claims. The Summary and Abstract sections may set forth one or more but not all exemplary embodiments of the present invention as contemplated by the inventor(s), and thus, are not intended to limit the present invention and the appended claims in any way. While specific configurations and arrangements are discussed, it should be understood that this is done for illustrative purposes only. A person skilled in the pertinent art will recognize that other configurations and arrangements can be used without departing from the spirit and scope of the present invention. It will be apparent to a person skilled in the pertinent art that this invention can also be used in a variety of other applications. The scope of the invention is not limited to the disclosed embodiments. The invention is defined by the claims appended hereto. References to “one embodiment,” “an embodiment,” “this embodiment,” “an example embodiment,” etc., indicate that the embodiment described may include a particular feature, structure, or characteristic, but every embodiment might not necessarily include the particular feature, structure or characteristic. Moreover, such phrases are not necessarily referring to the same embodiment. Further, when a particular feature, structure, or characteristic is described in connection with an embodiment, it is understood that it is within the knowledge of one skilled in the art to effect such a feature, structure, or characteristic in connection with other embodiments whether or not explicitly described. In some instances, one or more components may be referred to herein as “configured to,” “configurable to,” “operable/operative to,” “adapted/adaptable,” “able to,” “conformable/conformed to,” etc. Those skilled in the art will recognize that such terms (e.g. “configured to”) can generally encompass active-state components and/or inactive-state components and/or standby-state components, unless context requires otherwise. Those having skill in the art will recognize that the state of the art has progressed to the point where there is little distinction left between hardware, software, and/or firmware implementations of aspects of systems; the use of hardware, software, and/or firmware is generally (but not always, in that in certain contexts the choice between hardware and software can become significant) a design choice representing cost vs. efficiency tradeoffs. Those having skill in the art will appreciate that there are various vehicles by which processes and/or systems and/or other technologies described herein can be effected (e.g., hardware, software, and/or firmware), and that the preferred vehicle will vary with the context in which the processes and/or systems and/or other technologies are deployed. For example, if an implementer determines that speed and accuracy are paramount, the implementer may opt for a mainly hardware and/or firmware vehicle; alternatively, if flexibility is paramount, the implementer may opt for a mainly software implementation; or, yet again alternatively, the implementer may opt for some combination of hardware, software, and/or firmware. Hence, there are several possible vehicles by which the processes and/or devices and/or other technologies described herein may be effected, none of which is inherently superior to the other in that any vehicle to be utilized is a choice dependent upon the context in which the vehicle will be deployed and the specific concerns (e.g., speed, flexibility, or predictability) of the implementer, any of which may vary. Those skilled in the art will recognize that optical aspects of implementations will typically employ optically-oriented hardware, software, and or firmware. The foregoing detailed description has set forth various embodiments of devices and/or processes via the use of block diagrams, flowcharts, and/or examples. Insofar as such block diagrams, flowcharts, and/or examples contain one or more functions and/or operations, it will be understood by those within the art that each function and/or operation within such block diagrams, flowcharts, or examples can be implemented, individually and/or collectively, by a wide range of hardware, software, firmware, or virtually any combination thereof. In an embodiment, several portions of the subject matter described herein may be implemented via Application Specific Integrated Circuits (ASICs), Field Programmable Gate Arrays (FPGAs), digital signal processors (DSPs), or other integrated formats. However, those skilled in the art will recognize that some aspects of the embodiments disclosed herein, in whole or in part, can be equivalently implemented in integrated circuits, as one or more computer programs running on one or more computers (e.g., as one or more programs running on one or more computer systems), as one or more programs running on one or more processors (e.g., as one or more programs running on one or more microprocessors), as firmware, or as virtually any combination thereof, and that designing the circuitry and/or writing the code for the software and or firmware would be well within the skill of one of skill in the art in light of this disclosure. In addition, those skilled in the art will appreciate that the mechanisms of the subject matter described herein are capable of being distributed as a program product in a variety of forms, and that an illustrative embodiment of the subject matter described herein applies regardless of the particular type of signal bearing medium used to actually carry out the distribution. Examples of a signal bearing medium include, but are not limited to, the following: a recordable type medium such as a floppy disk, a hard disk drive, a Compact Disc (CD), a Digital Video Disk (DVD), a digital tape, a computer memory, etc.; and a transmission type medium such as a digital and/or an analog communication medium (e.g., a fiber optic cable, a waveguide, a wired communications link, a wireless communication link (e.g., transmitter, receiver, transmission logic, reception logic, etc.), etc.). By way of overview, illustrative embodiments provide nuclear fission reactors, and apparatuses and methods for their operation and simulation. Illustrative embodiments and aspects include, without limitation, nuclear fission reactors and reactor modules, including modular nuclear fission reactors and reactor modules, nuclear fission deflagration wave reactors and reactor modules, modular nuclear fission deflagration wave reactors and modules, methods of operating nuclear reactors and modules including the aforementioned, methods of simulating operating nuclear reactors and modules including the aforementioned, and the like. Still by way of overview and referring to FIG. 1A, an illustrative nuclear fission reactor 10 will be discussed by way of illustration and not limitation. Nuclear fission reactor 10 may be, but is not limited to, a fission deflagration wave reactor. A reactor 10 suitably includes a nuclear reactor core 100 disposed within a reactor vessel 12 and a reactor coolant system having one or more reactor coolant loops 14. A reactor may be a modular design including one or more nuclear reactor modules—see, e.g., an exemplary modular reactor 50 illustrated in FIG. 1B. Each reactor module 12 may be operatively coupled in fluid communication to at least one heat sink 58 via a reactor coolant system 56. Thus, each of the nuclear reactor modules may be considered a complete, stand-alone nuclear reactor system by itself. A nuclear reactor module may be neutronically coupled to at least one other adjacent reactor module. Thus, adjacent nuclear reactor modules can be neutronically integrated yet physically separate from each other. In order to provide an understanding of the control and simulation of reactors such as reactor 10 and reactor 50, illustrative core nucleonics, given by way of non-limiting examples, will be set forth first. While many reactor embodiments are contemplated, several of these non-limiting examples are illustrated in U.S. patent application Ser. No. 12/069,907 entitled MODULAR NUCLEAR FISSION REACTOR, naming AHLFELD, CHARLES E., GILLELAND, JOHN ROGERS, HYDE, RODERICK A., ISHIKAWA, MURIEL Y., MCALEES, DAVID G., MYHRVOLD, NATHAN P., WHITMER, CHARLES, and WOOD, LOWELL L. as inventors, filed 12 Feb. 2008, U.S. patent application Ser. No. 11/605,943, entitled AUTOMATED NUCLEAR POWER REACTOR FOR LONG-TERM OPERATION, naming RODERICK A. HYDE, MURIEL Y. ISHIKAWA, NATHAN P. MYHRVOLD, AND LOWELL L. WOOD, JR. as inventors, filed 28 Nov. 2006, U.S. patent application Ser. No. 11/605,848, entitled METHOD AND SYSTEM FOR PROVIDING FUEL IN A NUCLEAR REACTOR, naming RODERICK A. HYDE, MURIEL Y. ISHIKAWA, NATHAN P. MYHRVOLD, AND LOWELL L. WOOD, JR. as inventors, filed 28 Nov. 2006, and U.S. patent application Ser. No. 11/605,933, entitled CONTROLLABLE LONG TERM OPERATION OF A NUCLEAR REACTOR, naming RODERICK A. HYDE, MURIEL Y. ISHIKAWA, NATHAN P. MYHRVOLD, AND LOWELL L. WOOD, JR. as inventors, filed 28 Nov. 2006, the entire contents of which are hereby incorporated by reference in their entireties. Then, details will be set forth regarding several illustrative embodiments and aspects of reactors. Considerations Before discussing details of the reactors such as reactor 10 and reactor 50, some considerations behind reactor embodiments will be given by way of overview but are not to be interpreted as limitations. Some reactor embodiments address many of the considerations discussed below. On the other hand, some other reactor embodiments may address one or a select few of these considerations, and need not accommodate all of the considerations discussed below. Certain of the nuclear fission fuels envisioned for use in reactor embodiments are typically widely available, such as without limitation uranium (natural, depleted, or enriched), thorium, plutonium, or even previously-burned nuclear fission fuel assemblies. Other, less widely available nuclear fission fuels, such as without limitation other actinide elements or isotopes thereof may be used in embodiments of the reactor. While some reactor embodiments contemplate long-term operation at full power, or some portion thereof, on the order of around ⅓ century to around ½ century or longer, an aspect of some reactor embodiments does not contemplate nuclear refueling. Other reactor embodiments contemplate nuclear refueling, however. In some cases, embodiments may contemplate burial in-place at end-of-life. Nuclear refueling may occur during shutdown periods and/or operation at power. It is also contemplated that nuclear fission fuel reprocessing may be avoided in some cases, thereby mitigating possibilities for diversion to military uses and other issues. Some reactor embodiments may be sited underground, thereby addressing large, abrupt releases and small, steady-state releases of radioactivity into the biosphere. Some embodiments may entail minimizing operator controls, thereby automating those embodiments as much as practicable. In some embodiments, a life-cycle-oriented design is contemplated, wherein those embodiments can operate from startup to shutdown at end-of-life. In some life-cycle oriented designs, the embodiments may operate in a substantially fully-automatic manner. Some embodiments lend themselves to modularized construction. Finally, some embodiments may be designed according to high power density or to selected power densities corresponding to a variety of design considerations, such as burn-up criteria, power demand, neutronic flux considerations, and other parameters. During operation, the materials (e.g., elements and isotopes of elements) in a reactor, especially a reactor core region, change over time. For example, fuel atoms fission into fission products. Atoms of fuel, structural materials, poisons (fission product poisons or poisons intentionally inserted into the reactor), and so forth may absorb neutrons and become other isotopes or elements. These changes may be accounted for by design and reactor control in both the short term and the long term. An ability to move materials throughout the core may increase a reactor's effective lifetime. Some features of various reactor embodiments result from some of the above considerations. For example, simultaneously accommodating desires to achieve ⅓-½ century (or longer) of operations at full power without shutdown for nuclear refueling and to avoid nuclear fission fuel reprocessing may entail use of a fast neutron spectrum. As another example, in some embodiments a negative temperature coefficient of reactivity (αT) is engineered-in to the reactor, such as via negative feedback on local reactivity implemented with strong absorbers of neutrons or other approaches to reactivity control. In the alternative or in addition, some embodiments are configured to control the fission process in whole or in part by achieving a spectral shift in a neutron flux using spectral control methods such as displacing and/or inserting a neutron moderator for some time period. As a further example, in some modular deflagration wave embodiments, a distributed thermostat enables a propagating nuclear fission deflagration wave mode of nuclear fission fuel burn. This mode simultaneously permits a high average burn-up of non-enriched actinide fuels, such as natural uranium or thorium, and use of a comparatively small “nuclear fission igniter” region of moderate isotopic enrichment of nuclear fissionable materials in the core's fuel charge. As another example, in some embodiments, multiple redundancy is provided in primary and secondary core cooling. Exemplary Embodiments of Nuclear Fission Reactors Now that some of the considerations behind some of the reactor embodiments have been set forth, further details regarding an exemplary embodiment of nuclear fission reactors will be explained. It is emphasized that the following description of exemplary nuclear reactor embodiments is given by way of non-limiting examples only and not by way of limitation. As mentioned above, several embodiments of reactors are contemplated, as well as further aspects of reactor 10. After details regarding an exemplary embodiment of reactor 10 are discussed, other embodiments and aspects will also be discussed. Still referring to FIG. 1A, an exemplary embodiment of reactor 10 includes a reactor core assembly 100 that is disposed within a reactor pressure vessel 12. Several embodiments and aspects of reactor core assembly 100 are contemplated that will be discussed later. Some of the features that will be discussed later in detail include nuclear fission fuel materials and their respective nucleonics, fuel assemblies, fuel geometries, and the operation and simulation of reactor core assembly 100 in a complete reactor system. The reactor pressure vessel 12 suitably is any acceptable pressure vessel known in the art and may be made from any materials acceptable for use in reactor pressure vessels, such as without limitation stainless steel. Within the reactor pressure vessel 12, a neutron reflector (not shown) and a radiation shield (not shown) surround reactor core assembly 100. In some embodiments, the reactor pressure vessel 12 is sited underground. In such cases, the reactor pressure vessel 12 can also function as a burial cask for reactor core assembly 100. In these embodiments, the reactor pressure vessel 12 suitably is surrounded by a region (not shown) of isolation material, such as dry sand, for long-term environmental isolation. The region (not shown) of isolation material may have a size of around 100 meters in diameter or so. However, in other embodiments, the reactor pressure vessel 12 is sited on or toward the Earth's surface. Reactor coolant loops 14 transfer heat from nuclear fission in reactor core assembly 100 to application heat exchangers 16. The reactor coolant may be selected as desired for a particular application. In some embodiments, the reactor coolant suitably is helium (He) gas. In other embodiments, the reactor coolant suitably may be other pressurized inert gases, such as neon, argon, krypton, xenon, or other fluids such as water or gaseous or superfluidic carbon dioxide, or liquid metals, such as sodium or lead, or metal alloys, such as Pb—Bi, or organic coolants, such as polyphenyls, or fluorocarbons. The reactor coolant loops suitably may be made from tantalum (Ta), tungsten (W), aluminum (Al), steel or other ferrous or non-iron groups alloys or titanium or zirconium-based alloys, or from other metals and alloys, or from other structural materials or composites, as desired. In some embodiments, the application heat exchangers 16 may be steam generators that generate steam that is provided as a prime mover for rotating machinery, such as electrical turbine-generators 18 within an electrical generating station 20. In such a case, reactor core assembly 100 suitably operates at a high operating pressure and temperature, such as above 1,000K or so and the steam generated in the steam generator may be superheated steam. In other embodiments, the application heat exchanger 16 may be any steam generator that generates steam at lower pressures and temperatures (that is, need not be superheated steam) and reactor core assembly 100 operates at temperatures less than around 550K. In these cases, the application heat exchangers 16 may provide process heat for applications such as desalination plants for seawater or for processing biomass by distillation into ethanol, or the like. Optional reactor coolant pumps 22 circulate reactor coolant through reactor core assembly 100 and the application heat exchangers 16. Note that although the illustrative embodiment shows pumps and gravitationally driven circulation, other approaches may not utilize pumps, or circulatory structures or be otherwise similarly geometrically limited. The reactor coolant pumps 22 suitably may be provided when reactor core assembly 100 is sited approximately vertically coplanar with the application heat exchangers 16, such that thermal driving head is not generated. The reactor coolant pumps 22 may also be provided when reactor core assembly 100 is sited underground. However, when reactor core assembly 100 is sited underground or in any fashion so reactor core assembly 100 is vertically spaced below the application heat exchangers 16, thermal driving head may be developed between the reactor coolant exiting the reactor pressure vessel 12 and the reactor coolant exiting the application heat exchangers 16 at a lower temperature than the reactor coolant exiting the reactor pressure vessel 12. When sufficient thermal driving head exists, reactor coolant pumps 22 need not be provided to provide sufficient circulation of reactor coolant through reactor core assembly 100 to remove heat from fission during operation at power. In some embodiments more than one reactor coolant loop 14 may be provided, thereby providing redundancy in the event of a casualty, such as a loss of coolant accident, a loss of flow accident, a primary-to-secondary leak or the like, to any one of the other reactor coolant loops 14. Each reactor coolant loop 14 may be rated for full-power operation, though some applications may remove this constraint. In some embodiments, closures 24, such as reactor coolant shutoff valves, are provided in lines of the reactor coolant system 14. In each reactor coolant loop 14, a closure 24 may be provided in an outlet line from the reactor pressure vessel 12 and in a return line to the reactor pressure vessel 12 from an outlet of the application heat exchanger 16. Closures 24 may be fast-acting closures that shut quickly under emergency conditions, such as detection of significant fission-product entrainment in reactor coolant. Closures 24 may be provided in addition to a redundant system of automatically-actuated valves (not shown). One or more heat-dump heat exchangers 26 are provided for removal of after-life heat (decay heat). Heat-dump heat exchanger 26 includes a primary loop that is configured to circulate decay heat removal coolant through reactor core assembly 100. Heat-dump heat exchanger 26 includes a secondary loop that is coupled to an engineered heat-dump heat pipe network (not shown). In some situations, for example, for redundancy purposes, more than one heat-dump heat exchanger 26 may be provided. Heat-dump heat exchanger 26 may be sited at a vertical distance above reactor core assembly 100 so sufficient thermal driving head is provided to enable natural flow of decay heat removal coolant without need for decay heat removal coolant pumps. However, in some embodiments decay heat removal pumps (not shown) may be provided. Reactor coolant pumps may be used for decay heat removal, where appropriate. Now that an overview of an exemplary embodiment of the reactor 10 has been given, other embodiments and aspects will be discussed. First, embodiments and aspects of reactor core assembly 100 will be discussed. An overview of reactor core assembly 100 and its nucleonics will be set forth first, followed by descriptions of exemplary embodiments and other aspects of reactor core assembly 100. Given by way of overview and in general terms, structural components of reactor core assembly 100 may be made of tantalum (Ta), tungsten (W), rhenium (Re), various alloys including but not limited to steels such as martensitic stainless steels (e.g., HT9), austenitic stainless steels (e.g., Type 316), or carbon composite, ceramics, or the like. These materials are suitable because of the high temperatures at which reactor core assembly 100 operates, and because of their creep resistance over the envisioned lifetime of full power operation, mechanical workability, and corrosion resistance. Structural components can be made from single materials, or from combinations of materials (e.g., coatings, alloys, multilayers, composites, and the like). In some embodiments, reactor core assembly 100 operates at sufficiently lower temperatures so that other materials, such as aluminum (Al), steel, titanium (Ti) or the like can be used, alone or in combinations, for structural components. In deflagration wave embodiments, reactor core assembly 100 may include a small nuclear fission igniter and a larger nuclear fission deflagration burn-wave-propagating region. The nuclear fission deflagration burn-wave-propagating region suitably contains thorium or uranium fuel, and functions on the general principle of fast neutron spectrum fission breeding. In some deflagration wave embodiments, uniform temperature throughout reactor core assembly 100 is maintained by thermostating modules which regulate local neutron flux and thereby control local power production. Some example deflagration wave embodiments are further discussed in the aforementioned U.S. patent application Ser. No. 11/605,933, entitled CONTROLLABLE LONG TERM OPERATION OF A NUCLEAR REACTOR (“the '933 application”), which is herein incorporated by reference in its entirety. Nuclear reactors may be modular. Referring now to FIG. 1B, an illustrative modular reactor 50 is shown. It is emphasized that the following description of an exemplary embodiment of reactor 50 is given by way of non-limiting example only and not by way of limitation. As mentioned above, several embodiments of reactors such as reactors 10 and 50, are contemplated, as well as further aspects of reactors. Features illustrated in reactors 10 and 50 may be implemented separately or in any suitable combination. After details regarding an exemplary embodiment of reactor 50 are discussed, other embodiments and aspects will also be discussed. Modular reactor 50 is shown by way of illustration and does not limit modular reactors to a toroidal arrangement or any other arrangement of reactor modules 52. It will be understood that no limitation to such a geometric arrangement or to any geometric arrangement of any type whatsoever is intended. To that end, additional arrangements of reactor modules 52 will be discussed further below. In the interest of brevity, the description of additional arrangements of reactor modules 52 is limited to those illustrated herein. However, it will be appreciated that reactor modules 52 may be arranged in any manner whatsoever as desired and may accommodates neutronic coupling of adjacent nuclear fission deflagration wave reactor modules 52. As discussed above, the exemplary modular reactor 50 suitably includes reactor modules 52. Each reactor module 52 may suitably include a reactor core 54 and a reactor coolant system 56. Each nuclear fission deflagration wave reactor module 52 may be operatively coupled in fluid communication to at least one heat sink 58 via one or more associated reactor coolant systems 56. That is, each reactor modules 52 suitably may be considered a complete, stand-alone nuclear reactor by itself. A reactor module 52 may be neutronically coupled to at least one adjacent reactor module 52. While many embodiments of the modular reactor 50 are contemplated, a common feature among many contemplated embodiments of modular reactor 50 is neutronic coupling of adjacent reactor modules 52 via origination of a nuclear fission deflagration wave, or “burnfront” as further discussed in the aforementioned U.S. patent application Ser. No. 12/069,907 entitled MODULAR NUCLEAR FISSION REACTOR (the '907 application”), which is herein incorporated by reference in its entirety. Referring now to FIG. 1C, heat energy can be extracted from a nuclear fission reactor core according to another embodiment. In a nuclear fission reactor 110, nuclear fission occurs in a heat generating region 120 (e.g., throughout the fuel-bearing core or propagated in a burning wavefront, for example). Heat absorbing material 160, such as a condensed phase density fluid (e.g., water, liquid metals, terphenyls, polyphenyls, fluorocarbons, FLIBE (2LiF—BeF2) and the like) flows through the region 120 as indicated by an arrow 150, and heat is transferred from the heat generating region 120 to heat absorbing material 160. In some embodiments, e.g., fast fission spectrum nuclear reactors, heat absorbing material 160 is chosen to be a nuclear inert material (such as He4) so as to minimally perturb the neutron spectrum. In other embodiments of nuclear fission reactor 110, the neutron content is sufficiently robust, so that a non-nuclear-inert heat absorbing material 160 may be acceptably utilized. Heat absorbing material 160 flows (e.g., by natural convection or by forced movement) to a heat extraction region 130 that is substantially out of thermal contact with heat generating region 120. Heat energy 140 is extracted from heat absorbing material 160 at heat extraction region 130. Heat absorbing material 110 can reside in either a liquid state, a multiphase state, or a substantially gaseous state upon extraction of the heat energy 140 in the heat extraction region 130. Exemplary Movements of Nuclear Reactor Materials Fuel materials include not only fuel materials, but also structural materials (e.g., cladding). Referring now to FIG. 2A, a reactor 200, which may include any type of fission reactor including those described elsewhere herein, may include nuclear fission fuel assemblies 210 disposed therein. The following discussion includes details of exemplary nuclear fission fuel assemblies 210 that may be used in reactor 200. Referring now to FIG. 2B and given by way of non-limiting example, in an embodiment the nuclear fission fuel assembly 210 suitably includes a nuclear fission fuel assembly 220. In an embodiment, nuclear fission fuel assembly 220 has been “previously burnt.” The term “previously burnt” means that at least some components of the nuclear fission fuel assembly have undergone neutron-mediated nuclear fission and that the isotopic composition of the nuclear fission fuel has been modified. That is, the nuclear fission fuel assembly has been put in a neutron spectrum or flux (either fast or slow), at least some components have undergone neutron-mediated nuclear fission and, as result, the isotopic composition of the nuclear fission fuel has been changed. Thus, a previously burnt nuclear fission fuel assembly 220 may have been previously burnt in any reactor including reactor 200, such as without limitation a light water reactor. Previously burnt fission fuel (e.g., in a previously burnt nuclear fission fuel assembly 220) may be chemically untreated subsequent to its previous burning. It is intended that nuclear fission fuel assembly 220 can include without limitation any type of nuclear fissionable material whatsoever appropriate for undergoing fission in a nuclear fission reactor, such as actinide or transuranic elements like natural thorium, natural uranium, enriched uranium, or the like. Nuclear fission fuel assembly 220 is clad with cladding 224. If nuclear fission fuel assembly has been previously burnt, the cladding 224 may be the “original” cladding in which the nuclear fission fuel assembly 220 was clad before it was burnt. In some other embodiments, a previously burnt nuclear fission fuel assembly 220 may not be clad with “original” cladding 224. For example, a previously burnt nuclear fission fuel assembly 220 may be retained in its original cladding 224, and a new cladding (not shown) may be disposed around an exterior of cladding 224. In some embodiments, the new cladding is made up of cladding sections (not shown) that are configured to help accommodate swelling into the void spaces. In other embodiments, the new cladding may be provided as a barrier, such as a tube, provided between an exterior of the cladding 224 and reactor coolant (not shown). Referring now to FIG. 3, an exemplary nuclear fission fuel structure 300 includes non-contiguous segments 320 of nuclear fission fuel material. Non-contiguous segments 320 may be in “neutronic” contact without being in physical contact. Nuclear fission fuel structure 300 may also include an optional nuclear fission igniter 310. As described in the aforementioned '933 application, nuclear fission igniter 310 may be used in deflagration propagating wave-type nuclear reactors. Referring now to FIG. 4, a modular nuclear fission fuel core 400 may include an optional neutron reflector/radiation shield 410 and modular assemblies 420. Modular assemblies 420 may be modular fuel assemblies having some fuel material content. Modular assemblies may also be modular poison assemblies (having some poison material content), modular structural assemblies (serving a primarily structural purpose), modular payload assemblies (designed to carry a payload of, for example, a material to be subjected to a neutron flux), modular blank assemblies (serving as a mere placeholder, for example, to reduce the nucleonic, flow, structural, and thermal perturbations induced by a void or void filled with coolant and/or moderator), or any combination of the above. Modular assemblies 420 are placed as desired within the assembly receptacles 430. Modular nuclear fission fuel core 400 may be operated in any number of ways. For example, all of the assembly receptacles 430 in the modular nuclear fission fuel core 400 may be fully populated with modular fuel assemblies 420 prior to initial operation. For example, in deflagration propagating wave-type nuclear reactor embodiments, prior to initial operation means prior to origination and propagation of a nuclear fission deflagration propagating wave burnfront within and through the modular fuel assemblies 420. In other reactor embodiments, prior to initial operation means prior to initial criticality or prior to a modular nuclear fission fuel core being exposed to a neutron flux. As another example, modular assemblies 420 may be removed from their respective assembly receptacles 430 and replaced with other modular assemblies 440 (of the same or different type), as desired; this emplacement is indicated by the arrow 444. For example, “burnt” fuel assemblies may be replaced with “unburnt” fuel assemblies, poison assemblies may be replaced with fuel assemblies, and so forth. The other modular nuclear assemblies 440 may be unused or may have previously been used. For example, in deflagration propagating wave-type nuclear reactor embodiments, modular fission fuel assemblies 420 may be removed and replaced with other modular nuclear fission fuel assemblies 440 after a nuclear fission deflagration wave burnfront has completely propagated through modular nuclear fission fuel assemblies 420. In other embodiments, modular assemblies 420 may be removed and replaced with other modular assemblies 440 for any reason (e.g., testing or experimental uses, redistribution of fuel or poisons, etc.). Such replacement strategies may be used to extend operation of modular nuclear fission fuel core 400 as desired. As another example, the modular nuclear fission fuel core 400 need not be fully populated with modular assemblies 420 prior to initial operation. For example, less than all of the assembly receptacles 430 can be populated with modular assemblies 420. In such a case, the number of modular fuel assemblies that are placed within the modular nuclear fission fuel core 400 can be determined based upon many reasons, such as a number of modular fuel assemblies that are available, power demand (e.g., electrical loading in watts), that will be ultimately be placed upon the modular nuclear fission fuel core 400, etc. Thus, continued or extended operation of the modular nuclear fission fuel core 400 can be enabled without initially fueling the entire modular nuclear fission fuel core 400 with modular fuel assemblies. It will be appreciated that the concept of modularity can be extended. For example, in other embodiments, a modular nuclear fission reactor can be populated with any number of nuclear fission reactor cores in the same manner that the modular nuclear fission fuel core 400 can be populated with any number of modular assemblies 420. To that end, the modular nuclear fission reactor can be analogized to the modular nuclear fission fuel core 400 and nuclear fission reactor cores can be analogized to the modular nuclear fission fuel assemblies 420. The several contemplated modes of operation discussed above for the modular nuclear fission fuel core 400 thus apply by analogy to a modular nuclear fission reactor. Core materials not in a modular assembly may also be moved in a reactor core. It is well known in the art to control reactivity (and thus core average temperature in an operating reactor having a negative coefficient of reactivity) using control rods or other devices. In addition, other neutron modifying structures are contemplated in embodiment of the invention. For example, referring now to FIGS. 5A and 5B, neutron modifying structures 530 can position neutron modifying (e.g., absorbing, reflecting, moderating, etc.) substances in a reactor 500, including a propagating burnfront nuclear fission reactor 550, for a variety of purposes. In an embodiment, neutron modifying structures 530 insert neutron absorbers, such as without limitation Li-6, B-10, or Gd, into nuclear fission fuel. In another embodiment, neutron modifying structures 530 insert neutron moderators, such as without limitation hydrocarbons or Li-7, thereby modifying the neutron energy spectrum, and thereby changing the neutronic reactivity of nuclear fission fuel in the local region. In some situations in a reactor 500 (including a propagating burnfront nuclear fission reactor 550) an effect of the neutron moderators is associated with detailed changes in the neutron energy spectrum (e.g., hitting or missing cross-section resonances), while in other cases the effects are associated with lowering the mean neutron energy of the neutron environment (e.g., downshifting from “fast” neutron energies to epithermal or thermal neutron energies). In yet other situations, an effect of the neutron moderators is to deflect neutrons to or away from selected locations. In some embodiments, one of the aforementioned effects of neutron moderators is of primary importance, while in other embodiments, multiple effects are of comparable or lesser design significance. In another embodiment, neutron modifying structures 530 contain both neutron absorbers and neutron moderators; in one nonlimiting example, the location of neutron absorbing material relative to that of neutron moderating material is changed to affect control (e.g., by masking or unmasking absorbers, or by spectral-shifting to increase or decrease the absorption of absorbers), in another nonlimiting example, control is affected by changing the amounts of neutron absorbing material and/or neutron moderating material. In embodiments such as propagating burnfront nuclear fission reactor 550, a nuclear fission deflagration wave burnfront can be driven into areas of nuclear fission fuel as desired, thereby enabling a variable nuclear fission fuel burn-up. In propagating burnfront nuclear fission reactor 550, a nuclear fission deflagration wave burnfront 510 is initiated and propagated. Neutron modifying structures 530 can direct or move the burnfront 510 in directions indicated by arrows 520. In an embodiment, neutron modifying structures 530 insert neutron absorbers behind burnfront 510, thereby driving down or lowering neutronic reactivity of fuel that is presently being burned by burnfront 510 relative to neutronic reactivity of fuel ahead of burnfront 510, thereby speeding up the propagation rate of the nuclear fission deflagration wave. In another embodiment, neutron modifying structures 530 insert neutron absorbers into nuclear fission fuel ahead of burnfront 510, thereby slowing down the propagation of the nuclear fission deflagration wave. In other embodiments, neutron modifying structures 530 insert neutron absorbers into nuclear fission fuel within or to the side of the burnfront 510, thereby changing the effective size of the burnfront 510. In another embodiment, neutron modifying structures 530 insert neutron moderators, thereby modifying the neutron energy spectrum, and thereby changing the neutronic reactivity of nuclear fission fuel that is presently being burned by the burnfront 510 relative to neutronic reactivity of nuclear fission fuel ahead of or behind the burnfront 510. Thus, local neutronic reactivity in reactor 500, and burnfront 510 in propagating burnfront nuclear fission reactor 550, can be directed as desired according to selected local reaction rate or propagation parameters. For example, local reaction rate parameters can include fission rate, a heat generation density, cross-section dimensions of power density, or the like. In burnfront nuclear fission reactor 550, propagation parameters can include a propagation direction or orientation of the burnfront 510, a propagation rate of the burnfront 510, power demand parameters such the heat generation density, cross-sectional dimensions of a burning region through which the burnfront 510 is to the propagated (such as an axial or lateral dimension of the burning region relative to an axis of propagation of the burnfront 510), or the like. For example, the propagation parameters may be selected so as to control the spatial or temporal location of the burnfront 510, so as to avoid failed or malfunctioning control elements (e.g., neutron modifying structures or thermostats), or the like. Neutron modifying structures 530 may be actively controlled and/or passively controlled (e.g., programmable). Actively controlled neutron modifying structures are actively controlled by an operator and/or an external control system. Passively controlled neutron modifying structures are responsive to conditions at one or more locations in the core. For example, programmable temperature responsive neutron modifying structures (examples of which are discussed in detail in the aforementioned '933 application) introduce and remove neutron absorbing or neutron moderating material into and from the fuel-charge of a reactor 500 (including embodiments such as propagating burnfront nuclear fission reactor 550). Responsive to an operating temperature profile, programmable temperature responsive neutron modifying structures introduce neutron absorbing or moderating material into the fuel-charge of the nuclear fission reactor to lower operating temperature in the nuclear fission reactor or remove neutron absorbing or moderating material from the fuel-charge of the nuclear fission reactor in order to raise operating temperature of the nuclear fission reactor. It will be appreciated that temperatures are only one example of control parameters which can be used to determine the control settings of passively controlled or programmable neutron modifying structures. Nonlimiting examples of other control parameters which can be used to determine the control settings of programmable neutron modifying structures include power levels, neutron levels, neutron spectrum, neutron absorption, fuel burnup levels, and the like. In one example, the neutron modifying structures are used to control fuel burnup levels to relatively low (e.g., <50%) levels in order to achieve high-rate “breeding” of nuclear fission fuel for use in other nuclear fission reactors, or to enhance suitability of the burnt nuclear fission fuel for subsequent re-propagation of a nuclear fission deflagration wave in a propagating nuclear fission deflagration wave reactor. Different control parameters can be used at different times, or in different portions of the reactor. It will be appreciated that the various neutron modifying methods discussed previously in the context of neutron modifying structures can also be utilized in programmable temperature responsive neutron modifying structures, including without limitation, the use of neutron absorbers, neutron moderators, combinations of neutron absorbers and/or neutron moderators, variable geometry neutron modifiers, and the like. A material may be subjected to a neutron flux in a reactor. It should be appreciated that the neutron irradiation of material in a reactor may be controlled by the duration and/or extent of duration and local power level. In another embodiment, the neutron irradiation of material may be controlled by control of the neutron environment (e.g., the neutron energy spectrum for Np-237 processing) via neutron modifying structures. Referring to FIGS. 6A and 6B, for example, a material 610 inserted into a reactor 600, as indicated generally with arrow 602, will be subject to a neutron flux dependent upon, inter alia, local power level, duration, and neutron modifying structures. In an embodiment where the reactor is a propagating nuclear fission deflagration wave reactor, such as reactor 650, material 610 may be inserted into reactor 650 as indicated generally with arrow 652. In another embodiment, propagating nuclear fission deflagration wave reactor 650 may be operated in a “safe” sub-critical manner, relying upon an external source of neutrons to sustain the propagating burnfront, while using a portion of the fission-generated neutrons for nuclear processing of core materials. It should be appreciated that the movement of material 610 to a location within reactor 600 (or 650) may be from a location external to the reactor (as shown) or from another location within the reactor (not shown). In some embodiments, a material 610 may be present in a location within the reactor before nuclear fission ignition occurs within a reactor, while in other embodiments the material may be added (i.e., moved to the location) after nuclear fission occurs or occurs in that locale. In some embodiments, material is removed from the reactor, while in other embodiments it remains in place. Alternately, a material having a set of non-irradiated properties is loaded into a reactor. The material is transported (e.g., as indicated generally by arrows 652 and 602) into physical proximity and neutronic coupling with a region of maximized reactivity—in the case of propagating nuclear fission deflagration wave reactor 650, as the nuclear fission deflagration wave propagating burnfront (e.g., burnfront 670) passes through the material. The material 610 remains in neutronic coupling for a sufficient time interval to convert the material 610 into a second material 606 having a desired set of modified properties. Upon the material 610 having thus been converted into the material 606, the material 606 may be physically transported out of reactor 600 (or reactor 650) as generally indicated by arrow 604 (or 654). The removal can take place either during operation of reactor 600 (or 650) or after shutdown. The removal can be performed as a continuous, sequential, or batch process. In one example, nuclearly processed material 606 may be subsequently used as nuclear fission fuel in another nuclear fission reactor, such as without limitation LWRs or propagating nuclear fission deflagration wave reactors. In another nonlimiting example, nuclearly processed material 606 may be subsequently used within the nuclear fission ignitor of a propagating nuclear fission deflagration wave reactor. In one approach, thermal management may be adjusted to provide thermal control appropriate for any changes in operational parameters, as appropriate for the revised materials or structures. According to further embodiments, temperature-driven neutron absorption can be used to control a nuclear fission reactor, thereby “engineering-in” an inherently-stable negative temperature coefficient of reactivity (αT). Referring now to FIG. 7A, a nuclear reactor 700 is instrumented with temperature detectors 710, such as without limitation thermocouples. In this embodiment, the nuclear fission reactor 700 suitably can be any type of fission reactor whatsoever. To that end, the nuclear fission reactor 700 can be a thermal neutron spectrum nuclear fission reactor or a fast neutron spectrum nuclear fission reactor, as desired for a particular application. For example, temperature detectors detect local temperature in reactor 700 and generate a signal 714 indicative of a detected local temperature. The signal 714 is transmitted to a control system 720 in any acceptable manner, such as without limitation, fluid coupling, electrical coupling, optical coupling, radiofrequency transmission, acoustic coupling, magnetic coupling, or the like. Responsive to signal 714 indicative of the detected local temperature, control system 720 determines an appropriate correction (positive or negative) to a local neutronic reactivity of reactor 700 (e.g., to return reactor 700 to a desired operating parameter, such as desired local temperatures during reactor operations at power). To that end, control system 720 generates a control signal 724 indicative of a desired correction to local neutronic reactivity. Control signal 724 is transmitted to a dispenser 730 of neutron absorbing material. Control signal 724 suitably may be transmitted in the same manner or a different manner as signal 714. The neutron absorbing material suitably may be any neutron absorbing material as desired for a particular application, such as without limitation Li-6, B-10, or Gd. Dispenser 730 suitably is dispensing mechanism acceptable for a desired application. A reservoir (not shown) may be located locally to dispenser 730 or may be located remotely from the dispensing mechanism 730 (e.g., outside a neutron reflector of reactor 700). Dispenser 730 dispenses the neutron absorbing material within the nuclear fission reactor core responsive to the control signal 1124, thereby altering the local neutronic reactivity. Referring now to FIG. 7B and given by way of non-limiting example, exemplary thermal control may be established with a neutron absorbing fluid. A thermally coupled fluid containing structure 740 contains a fluid in thermal communication with a local region of reactor 700. The fluid in the structure 740 expands or contracts responsive to local temperature fluctuations. Expansion and/or contraction of the fluid is operatively communicated to a force coupling structure 750, such as without limitation a piston, located external to the nuclear fission reactor 700. A resultant force communicated by the force coupling structure 750 is exerted on neutron absorbing fluid in a neutron absorbing fluid containing structure 760. The neutron absorbing fluid is dispensed accordingly from the structure 760, thereby altering the local neutronic reactivity. In another example, a neutron moderating fluid may be used instead of, or in addition to, the neutron absorbing fluid. The neutron moderating fluid changes the neutron energy spectrum and lowers the mean neutron energy of the local neutron environment, thereby driving down or lowering neutronic reactivity of nuclear fission fuel within the nuclear fission reactor 700. In another example, the neutron absorbing fluid and/or the neutron modifying fluid may have a multiple phase composition (e.g., solid pellets within a liquid). FIG. 7C illustrates details of an exemplary implementation of the arrangement shown in FIG. 7B. Referring now to FIG. 7C, fuel power density in a nuclear fission reactor 701 is continuously regulated by the collective action of a distributed set of independently-acting thermostating modules, over very large variations in neutron flux, significant variations in neutron spectrum, large changes in fuel composition and order-of-magnitude changes in power demand on the reactor. This action provides a large negative temperature coefficient of reactivity just above the design-temperature of reactor 701. Located throughout the fuel-charge in the nuclear fission reactor 701 in a 3-D lattice (which can form either a uniform or a non-uniform array) whose local spacing may be roughly a mean free path of a median-energy-for-fission neutron (it may be reduced for redundancy purposes), each of these modules includes a pair of compartments, each one of which is fed by a capillary tube. A small thermostat-bulb compartment 761 located in the nuclear fission fuel contains a thermally sensitive material, such as without limitation, Li-7, whose neutron absorption cross-section may be low for neutron energies of interest, while the relatively large compartment 741 is positioned in a different location (e.g., on the wall of a coolant tube) and may contain variable amounts of a neutron absorbing material, such as without limitation, Li-6, which has a comparatively large neutron absorption cross-section. At a pressure of 1 bar, lithium melts at 453K and boils at 1615K, and therefore is a liquid across typical operating temperature ranges of reactor 701. As fuel temperature rises, the thermally sensitive material contained in the thermostat-bulb 761 expands, and a small fraction of it is expelled (e.g., approximately 10−3, for a 100K temperature change in Li-7), potentially under kilobar pressure, into the capillary tube which terminates on the bottom of a cylinder-and-piston assembly 751 located remotely (e.g., outside of the radiation shield) and physically lower than the neutron absorbing material's intra-core compartment 741 (in the event that gravitational forces are to be utilized). There the modest volume of high-pressure thermally sensitive material drives a swept-volume-multiplying piston in the assembly 751 which pushes a larger (e.g., potentially three order-of-magnitude larger) volume of neutron absorbing material through a core-threading capillary tube into an intra-core compartment proximate to the thermostat-bulb which is driving the flow. There the neutron absorbing material, whose spatial configuration is immaterial as long as its smallest dimension is less than a neutron mean free path, acts to absorptively depress the local neutron flux, thereby reducing the local fuel power density. When the local fuel temperature drops, neutron absorbing material returns to the cylinder-and-piston assembly 751 (e.g., under action of a gravitational pressure-head), thereby returning the thermally sensitive material to the thermostat-bulb 761 whose now-lower thermomechanical pressure permits it to be received. It will be appreciated that operation of thermostating modules does not rely upon the specific fluids (Li-6 and Li-7) discussed in the above exemplary implementation. In one exemplary embodiment, the thermally sensitive material may be chemically, not just isotopically, different from the neutron absorbing material. In another exemplary embodiment, the thermally sensitive material may be isotopically the same as the neutron absorbing material, with the differential neutron absorbing properties due to a difference in volume of neutronically exposed material, not a difference in material composition. Reactor Control and Simulation The aforementioned examples thus demonstrate that fuel, poison, and other materials may be moved throughout a reactor core by several mechanisms with or without moving complete assemblies. Such movements may complicate calculations of nuclide concentrations (i.e., numbers of atoms and isotopes and nuclear isomers of atoms per unit volume) in the core. In general, the calculation of nuclide concentrations in the core or an operating reactor or simulation thereof may be broken into two interrelated parts: neutron transport and transmutation. Neutron transport calculations may determine neutron populations (e.g., flux and flux spectrum), while transmutation calculations determine the populations of nuclides given a starting population and a neutron flux. Neutron transport calculations can be done, for example, using deterministic methods (e.g., a discrete ordinates method), using stochastic methods such as a Monte Carlo method, or by using a hybrid of the two (e.g., using deterministic methods to calculate certain aspects in an otherwise Monte Carlo implementation). Deterministic methods typically solve transport equations using average particle behavior. A discrete method typically divides the phase space into many small volumes. Neutrons moving between adjacent volumes take a small amount of time to move a small distance. Thus, calculation approaches the integro-differential transport equation (having space and time derivatives) as time, volume, and distance are made smaller, i.e., approach 0. Monte Carlo methods, on the other hand, obtain answers by simulating individual particles and recording some aspects of their average behavior. Monte Carlo methods are often used when it is difficult to determine an example result using a deterministic method. As applied to neutron transport, a Monte Carlo method may simulate the individual probabilistic events, thus following neutrons through their lifecycle from birth to death (e.g., absorption, escape, etc.). The associated probability distributions (e.g., represented by continuous and/or discrete probability density functions) are randomly sampled to determine the outcome (e.g., scatter, fission, neutron capture, leakage) at each time step. Collisions may be modeled using physics equations and cross sectional data. The frequency of collisions, and thus neutron induced reactions such as fission and loss due to absorption by poisons are, of course, dependent on the concentration of fissile isotopes and poisons respectively in the volume of interest. Cross-sectional data for an atom represents the effective cross sectional area that an atom presents to a particle for an interaction, e.g., for a neutron, for interactions such as the various scattering and absorption types. Cross sections typically vary by the atom, the particle, and the energy of the particle. Thus, a cross section may be used to express the likelihood of a particular interaction of an atom with an incident particle having a certain energy. Microscopic properties, such as a microscopic cross section for a reaction (e.g., scatter, radiative capture, absorption, fission), are intrinsic properties of a type of nuclei (i.e., of a specific material's nuclei). Macroscopic properties, such as a macroscopic cross section for a reaction, is a property of a volume of the material having a concentration or density (e.g., in number of atoms per unit volume) of the material. Microscopic cross section is typically expressed in units of area (e.g., cm2 or “barns”—a barn is 10−28 m2). Macroscopic cross sections are proportional to the microscopic cross section multiplied by the density, or equivalently 1/(mean free path length) and thus are expressed in units of 1/length (e.g., m−1). Cross sectional data is typically determined by empirical means. Thus, especially for short-lived isotopes, cross sectional data for a large spectrum of neutron energies is simply not available yet. Performing accurate Monte Carlo calculations on volumes having a population of isotopes not having completely known or well-characterized properties such as neutron cross-sections is therefore difficult. Additionally, even if all the cross sectional data for each and every material was well characterized, the computational burden would be significant. Methods which may help reduce these difficulties and/or computational burdens are described in detail elsewhere herein. Transmutation calculations determine the inventory or concentration of each nuclide as it varies, for example, under a neutron flux. In general, transmutation calculations may be thought of as determining a new population of a material based on the loss rate and the production rate of the material subject to a given neutron flux. A given atom of a material may, for example, fission and produce two fission products; while another atom of the material might be converted to an isotope of a larger atomic mass number (A) after capturing a neutron. Yet another atom of the material might beta or alpha decay to another element, and so forth. Thus, the rate of change of an amount of a material in an operating reactor is typically the sum of the loss rate due to decay, gain rate due to decay, loss due to neutron-induced reactions, and gain due to neutron-induced reactions. It is to be appreciated that transmutation calculations for materials depend upon the current neutron flux, and neutron flux calculations depend upon the current concentration of materials such as fissile isotopes and poisons. These calculations may be linked together in various ways, including but not limited to such iterative numerical analysis tools such as the Runge-Kutta methods. A complete description of Runge-Kutta is not necessary, as it is well known in the art. In general, however, explicit Runge-Kutta methods, “solve” the initial value problemy′=f(t,y),y(t0)=y0 using the equations y n + 1 = y n + h ∑ i = 1 s b i k i where k1=f(tn, yn), k2=f(tn+c2h, yn+a21hk1), k3=f(tn+c3h, yn+a31hk1+a32hk2), ks=f(tn+csh, yn+as1hk1+as2hk2+ . . . +as,s-1hks-1) To specify a specific Runge Kutta method, one may supply an integer, s, and a set of coefficients aij, bij, and ci, The Runge Kutta method is consistent if the coefficients are such that: ∑ j = 1 i - 1 a ij = c i for i = 2 , … , s . Thus, for example, a consistent fourth order Runge Kutta is:yn+1=yn+⅙h(k1+k2+k3+k4),tn+1=tn+h where k1=f(tn, yn), k2=f(tn+½h,yn+½hk1), k3=f(tn+½ h,yn+½hk2), and k4=f(tn+h,yn+hk3). Thus, the next value, yn+1, is determined by the present value, yn, plus the product of the size of the interval and an estimated slope. The slope is a weighted average of slopes: k1 is the slope at the beginning of the interval, k2 is the slope at the midpoint of the interval using slope k1 to determine the value of y at the point tn+h/2 using Euler's method; k3 is again the slope at the midpoint, but now using the slope k2 to determine the y-value; and k4 is the slope at the end of the interval, with its y-value determined using k3. The Euler method is a one stage Runge Kutta method. The Euler method essentially estimates the slope and advances a small step using that slope. Examples of second order Runge Kutta methods include the midpoint method and Heun's method. Thus, an updated amount (e.g., inventory or concentration) of a material in a reactor core or volume of interest (inside or outside the reactor core) may be determined by determining an average rate of change of the amount of the material based on the previous amount of the material and a neutron flux. This may be performed individually or simultaneously for all of the materials in the reactor core or the volume of interest. The neutron flux, in turn, may be determined by determining an average rate of change of flux based on the amount of the materials in the core. Accuracy of the calculations may be enhanced if subvolumes of a reactor are considered rather than a reactor core in gross. For example, gross calculations may be performed on a homogenous model of a reactor core—the core is simulated to have an even distribution of all materials. Higher resolution may be obtained by representing the core as a volume comprised of many homogeneous cells, each cell being allowed to have different concentrations of materials. Although cells need not be homogenous, homogenous cells are typically preferred to simplify calculations. If the resolution is high enough, the core may be represented with very good precision. For example, a three-dimensional geometry of cells, each having a defined geometry and concentrations of materials may be used. Cells may be defined in many ways, including but not limited to by their bounding surfaces such as equations of surfaces and intersections and unions of regions of space. Transport calculations typically determine for each cell the number of reactions and boundary crossings to each neighboring cell. As illustrated in FIG. 8A, a structure 800 may be formed by cells having complicated shapes. For the sake of simplicity, only two dimensions are shown (i.e., a cross section), but it is understood that cells are typically three dimensional. Moreover, in this non-limiting example, the locations and shapes are relatively uniform. For example, exemplary cell 802 may be a sphere. Exemplary cell 804 may be a larger sphere excluding the volume defined by cell 802. Exemplary cell 806 may be a cube, excluding the volume circumscribed by the outer spherical surface of cell 804. Alternatively, cell 802 could be a cylinder extending some distance into the figure, cell 804 could be the volume determined by a larger cylinder excluding the volume of cell 802, and cell 806 could be a rectangular prism excluding the volume within the cylinder defined by cell 804's outer surface. In any case, cell 802 may include one composition of fuel materials, poison materials, and structural materials. Cell 804 may have a second composition of fuel materials, poison materials, and structural materials. Cell 806 may be a third composition of structural materials only (e.g., cladding). As illustrated in FIG. 8B, cells may be combined to form larger structures. For example, structure 800 may represent a rectangular prism-shaped fuel assembly. Structure 830 includes many structures 800. For example, structure 830 may define a fuel module of six fuel pins by four fuel pins and fifty fuel pins deep. Thus, even larger structures may be formed. For example, as illustrated by FIG. 8C, exemplary structure 860 may represent a reactor core having an arrangement of nineteen structures 830 (e.g., fuel modules) each including many structures 800 (fuel assemblies). Thus, specific physical locations in space of an actual operating reactor or a detailed reactor design may be represented by a cell. Calculations may be performed using a detailed model representing an actual reactor during operation. The results may be used to make decisions regarding reactor control. Similarly, calculations may be performed on a representation of a proposed reactor to test operating procedures or to test proposed fuel and poison loading. Transmutation and transport calculations may be performed for each cell. For a complex model, this can result in a large computational burden due in part to the large number of cells. The computational burden is also increased by the number of materials which may be present in each cell. Prior to operation, a reactor already contains a large number of materials (e.g., various fuel isotopes, installed poison isotopes, structural isotopes, moderator, reflectors, etc.). Immediately upon operation, however, the number of materials (e.g., isotopes) in the reactor increases significantly due to neutron capture and especially neutron-induced fission. The distribution of fission products from a fission of a given isotope induced by a neutron of a given energy may be described by a fission product yield curve. FIG. 9 illustrates an exemplary fission product yield curve 900. It should be appreciated that the graph illustrates the total fission yield in percent of fission products having each mass number (A). More than one isotope may have a given mass number. Thus, fission products having a mass number of, for example, 140, fall under the point on the curve defined by mass number=140. In this example, the fission products produced by the thermal fission of U-235 is illustrated on fission product yield curve 900. Curves for fissions of U-235 induced by fast neutrons will have a similar but different shape. Neutron energies may be classified in more detail than “fast” or “thermal.” Also, fission yield curves for other fissile isotopes will have a similar but different shape. In general, however, fission yield curves follow this “M” shape having two peaked “humps.” Thus, the curve may be divided into two portions, left curve portion 912 which includes a left peak 922, and right curve portion 914 which includes a right peak 924. Thus area 902 falls under left peak 922 and left curve portion 912 and area 904 falls under right peak 924 and right curve portion 914. As a reactor operates, the level of fission products tends to increase due to fission (i.e., have a production rate due to fission), but tends to decrease due to decay and neutron capture or “burnout” (i.e., have loss rates due to decay and capture). Transmutation calculations may be used to determine or approximate these levels during reactor operation. As discussed elsewhere herein, reactor control systems, such as control system 720, may determine appropriate corrections (positive or negative) to a local neutronic reactivity of reactor 700 (e.g., to return reactor 700 to a desired operating parameter, such as desired local temperatures during reactor operations at power). To that end, control systems may generate a control signal (e.g., control signal 724) indicative of a desired correction to local neutronic reactivity. Reactor control systems and control signals are not limited to the embodiments such as control system 720 and control signal 724. Reactor Control Systems may also control other neutron affecting or absorbing features such as control rods, to control and/or shut down the reactor as desired, which is well known in the art. Reactor Control Systems may also generate control signals to order changes in various flows, e.g., the flow of heat absorbing material (e.g., coolant) through the reactor or portions of the reactor by ordering changes in reactor coolant pump (e.g., reactor coolant pumps 22) operation and/or various valve positions in the reactor system, including but not limited to reactor closures (e.g., closures 24) or reactor coolant shutoff valves, steam shutoff valves, etc. Reactor Control Systems may also order changes in breaker positions (e.g., reactor coolant pump power supply breakers, steam turbine-generator output breakers, etc.). As is well known in the art, Reactor Control Systems may have temperature inputs (e.g., control system 720 receiving input from temperature detectors 710) in addition to neutron detectors (e.g., to sense neutron flux to determine reactor power or local reactor power at a portion of the core), and flow and position detectors (e.g., venturi-type flow detectors, valve position indicators, breaker position indicators). Thus, Reactor Control Systems may control the flow of heat absorbing material (e.g., coolant) through the reactor and/or portions of the reactor to control overall temperatures and local temperatures in response to overall reactor thermal power and/or local reactor thermal power. Reactor Control Systems may also provide operator indications and accept operator inputs. Thus, a Reactor Control System monitors reactor operations, may provide some automatic control features (such as changing flow rates and moving control rods or otherwise positioning neutron affecting or absorbing materials, which are described in more detail elsewhere herein), displays operational parameters, and accepts and executes operator inputs for manual control actions. Example Computer System Some aspects and/or features of the present invention can be implemented by software, firmware, hardware, or a combination thereof. Calculations may be approximated using table look-ups. Hardware implementations of individual components are not limited to digital implementations and may be analog electrical circuits. Additionally, embodiments may be realized in a centralized fashion in at least one communication system, or in a distributed fashion where different elements may be spread across several interconnected communication systems. Any kind of computer system or other apparatus adapted for carrying out the methods described herein may be suited. FIG. 10 illustrates an example computer system 1000 in which the present invention, or portions thereof, can be implemented as computer-readable code. Various embodiments of the invention are described in terms of this example computer system 1000. After reading this description, it will become apparent to a person skilled in the relevant art how to implement the invention using other computer systems and/or computer architectures. Computer system 1000 includes one or more processors, such as processor 1004. Processor 1004 can be a special purpose or a general purpose processor. Processor 1004 is connected to a communication infrastructure 1006 (for example, a bus or network). Computer system 1000 also includes a main memory 1008, preferably random access memory (RAM), and may also include a secondary memory 1010. Secondary memory 1010 may include, for example, a hard disk drive 1012, a removable storage drive 1014, any type of non-volatile memory, and/or a memory stick. Removable storage drive 1014 may comprise a floppy disk drive, a magnetic tape drive, an optical disk drive, a flash memory, or the like. The removable storage drive 1014 reads from and/or writes to a removable storage unit 1018 in a well known manner. Removable storage unit 1018 may comprise a floppy disk, magnetic tape, optical disk, etc. which is read by and written to by removable storage drive 1014. As will be appreciated by persons skilled in the relevant art(s), removable storage unit 1018 includes a computer usable storage medium having stored therein computer software and/or data. In alternative implementations, secondary memory 1010 may include other similar means for allowing computer programs or other instructions to be loaded into computer system 1000. Such means may include, for example, a removable storage unit 1022 and an interface 1020. Examples of such means may include a program cartridge and cartridge interface (such as that found in video game devices), a removable memory chip (such as an EPROM, or PROM) and associated socket, and other removable storage units 1022 and interfaces 1020 which allow software and data to be transferred from the removable storage unit 1022 to computer system 1000. Computer system 1000 may also include a communications interface 1024. Communications interface 1024 allows software and data to be transferred between computer system 1000 and external devices. Communications interface 1024 may include a modem, a network interface (such as an Ethernet card), a communications port, a PCMCIA slot and card, or the like. Software and data transferred via communications interface 1024 are in the form of signals which may be electronic, electromagnetic, optical, or other signals capable of being received by communications interface 1024. These signals are provided to communications interface 1024 via a communications path 1026. Communications path 1026 carries signals and may be implemented using wire or cable, fiber optics, a phone line, a cellular phone link, an RF link or other communications channels. Computer system 1000 may also be coupled to a Reactor Control system 1030. Reactor Control System 1030 may be directly interfaced to the communications infrastructure 1006 as shown in the figure. Reactor Control System may also be interfaced via communications interface 1024 or communications interface 1024 and communications path 1026. In this document, the terms “computer program medium” and “computer usable medium” are used to generally refer to media such as removable storage unit 1018, removable storage unit 1022, and a hard disk installed in hard disk drive 1012. Signals stored elsewhere and carried over communications path 1026 can also embody the logic described herein. Computer program medium and computer usable medium can also refer to memories, such as main memory 1008 and secondary memory 1010, which can be memory semiconductors (e.g. DRAMs, etc.). These computer program products are means for providing software to computer system 1000. Computer programs (also called computer control logic) are stored in main memory 1008 and/or secondary memory 1010. Computer programs may also be received via communications interface 1024. Such computer programs, when executed, enable computer system 1000 to implement the present invention as discussed herein. In particular, the computer programs, when executed, enable processor 1004 to be used in the performance of processes of the present invention, such as the methods illustrated by the flowcharts described elsewhere herein. Accordingly, such computer programs represent controllers of the computer system 1000. Where the invention is implemented using software, the software may be stored in a computer program product and loaded into computer system 1000 using removable storage drive 1014, interface 1020, hard drive 1012 or communications interface 1024. The invention is also directed to computer program products comprising software stored on any computer useable medium. Computer programs or software in the present context means any expression, in any language, code or notation, of a set of instructions intended to cause a system having an information processing capability to perform a particular function either directly or after either or both of the following: a) conversion to another language, code or notation; b) reproduction in a different material form. Such software, when executed in one or more data processing device, causes a data processing device(s) to operate as described herein. Embodiments of the invention employ any computer useable or readable medium, known now or in the future. Examples of computer useable mediums include, but are not limited to, primary storage devices (e.g., any type of random access memory), secondary storage devices (e.g., hard drives, floppy disks, CD ROMS, ZIP disks, tapes, magnetic storage devices, optical storage devices, MEMS, nanotechnological storage device, etc.), and communication mediums (e.g., wired and wireless communications networks, local area networks, wide area networks, intranets, etc.). Methods for Mapping Reactor Materials Now that illustrative embodiments of nuclear reactors and reactor control and simulation have been discussed, illustrative methods associated therewith will now be discussed. Following are a series of flowcharts depicting implementations of processes. For ease of understanding, the flowcharts are organized such that the initial flowcharts present implementations via an overall “big picture” viewpoint and thereafter the following flowcharts present alternate implementations and/or expansions of the “big picture” flowcharts as either sub-steps or additional steps building on one or more earlier-presented flowcharts. Those having skill in the art will appreciate that the style of presentation utilized herein (e.g., beginning with a presentation of a flowchart(s) presenting an overall view and thereafter providing additions to and/or further details in subsequent flowcharts) generally allows for a rapid and easy understanding of the various process implementations. In addition, those skilled in the art will further appreciate that the style of presentation used herein also lends itself well to modular design paradigms. The blocks may be performed in any order or concurrently unless specified otherwise. Some embodiments of the present invention do not require the performance of each and every block, regardless whether the block or blocks is/are explicitly labeled or described as optional. Other embodiments require the repetition of one or more blocks, regardless whether the block is labeled or described as repeated. Referring now to FIG. 11, an illustrative method 1100 is provided for simulating and/or controlling a nuclear reactor. The method 1100 starts at a block 1105. At block 1105, a flux in a first cell is determined based on the amount(s) of at least one material in at least the first cell. The flux determination may be further based on the amounts of more than one material in the first cell and/or a previous flux in the first cell. Also, the flux determination may be further based on the amounts one or more materials in one or more other cells. For example, a flux may be determined by a transport calculation (e.g., solving neutron transport equations). A “flux” may be any flux (e.g., photon, alpha, beta, etc.), but is typically a neutron flux. The flux may be determined by numerical analysis methods using an average rate of change of the flux. The average rate of change of the flux may be a weighted average (e.g., as determined by a Runge Kutta method or any other method). The flux may be dependent upon the amount(s) of one or more materials in the first cell. The flux may be further dependent upon the amount(s) of one or more materials in one or more additional cells. An “amount” may be a mass or a number (e.g., number of atoms) or may be a density/concentration (e.g., mass or number of particles per unit volume). A cell represents a physical location or region in a nuclear reactor. The reactor may be, for example, real or simulated, currently operating, or under design. The reactor may be any type or sub-type of reactor, including light water reactor, heavy water reactor, pressurized water reactor, boiling water reactor, propagating nuclear fission deflagration wave reactor, etc. The reactor is typically represented by many homogeneous cells, but heterogeneous cells may be used. Each cell may have the same or different shape or volume as any other cell. A material may be one or more of any element, molecule, family of elements, family of molecules, isotope, family of isotopes, isomers of isotopes, fertile isotope(s), fission product(s), fission product poisons, etc. Materials are typically elements and isotopes of elements. Thus, U-235 and U-238 are thus typically two different materials. At block 1110, the average rate(s) of change of the one or more amount(s) of one or more material(s) in the first cell is determined based on previous amount(s) of the material(s) and a flux in the first cell. For example, average rates of change may be determined by a transmutation rate calculation. The average rate of change of the amount of one or more materials may be determined by numerical analysis methods using an average rate of change of the amount. The average rate of change of the flux may be a weighted average (e.g., as determined by a Runge Kutta method or any other method). The amount may be dependent upon the flux in the first cell. The average rates of change for the one or more materials may be solved individually or simultaneously (such as when coupled through transmutation equations). At block 1115, updated amount(s) in the first cell for the material(s) are determined based on the average rate(s) of change. For example, updated amounts may be determined by performing transmutation calculations. The updated amounts for the one or more materials may be solved individually or simultaneously (such as when coupled through transmutation equations). At block 1120, at least one move quantity is determined. A move quantity may be any quantity of one or more materials such as a quantity of a material that is desired to be moved into or out of a cell. In this block one or more move quantities may each apply to one or more materials in the first cell. A move quantity may be determined in response to one or more reactor parameters such as a flux or a fluence, a power level (local or overall), a temperature, etc. A reactor parameter may be compared to a threshold or set point for that parameter. This block may be repeated as suitable, e.g., for each of one or more materials in the first cell. At block 1125, the updated amount(s) in the first cell is (are) adjusted by the move quantity(-ies). One or more move quantities are each applied to the amounts of one or more materials in the first cell, thus increasing or decreasing each affected amount. A move amount of zero may be used to signify no change. In an embodiment, a material may be moved outside the reactor. In this case, blocks 1130 through 1145 may be skipped. At block 1130, a flux in a second cell is determined based on amount(s) of at least one material in the second cell. As discussed above, the flux determination may be further based on the amounts of more than one material in the second cell. Also, the flux determination may be further based on the amounts one or more materials in one or more other cells. At block 1135, average rate(s) of change of the amount(s) of the material(s) in the second cell based on previous amount(s) of the material(s) and a flux in the second cell is determined. At block 1140, updated amount(s) in the second cell for the material(s) is (are) determined based on the average rate(s) of change in the second cell. At block 1145, the updated amount(s) in the second cell is (are) adjusted by the move quantity(-ies). At block 1150, a control action for a nuclear reactor is determined. A control action may be a change (positive or negative) to a local neutronic reactivity of a reactor using any neutron affecting or absorbing features such as movement of neutron absorbing materials or fluids, control rods, etc.; a change in one or more various flows, e.g., the flow of heat absorbing material (e.g., coolant) through the reactor by ordering changes in reactor coolant pump operation and/or various valve positions in the reactor system, including but not limited to reactor closures or reactor coolant shutoff valves, steam shutoff valves, etc.; a change in one or more breaker positions (e.g., reactor coolant pump power supply breakers, steam turbine-generator output breakers, etc.); or the like. The determined control action may be displayed to a user. In an embodiment, this block is optional. At block 1155, a control action for the nuclear reactor is performed. This performance may be automatic or manual. In an embodiment, this block is optional. At block 1160, approximately the move quantity(-ies) of the material(s) is/are transferred to/from the first cell location in a reactor. In this block, an actual of amount of at least one substance corresponding to one or more of the at least one material is transferred from or to one location (i.e., from the first cell location or to the first cell location). This block may be performed in conjunction with block 1155 or separately. The transferred amount of a substance (i.e., approximately the move quantity of the corresponding material or materials) may, but is not required, to be associated with a component in the location represented by the cell (e.g., an assembly including fuel, poison, structural components, or any combination of these). In an embodiment such as a simulation or evaluation of a reactor design, this step is optional. The method stops at block 1160, but may continue to point A as indicated in other methods in other figures. Referring now to FIG. 12, an illustrative method 1200 is provided for simulating and/or controlling a nuclear reactor. The method 1200 starts at a block 1210. As illustrated by point A, method 1200 may be preceded by method 1100. At block 1210, approximately the move quantity(-ies) of the material(s) is/are transferred from/to a second cell. In this block, an actual amount of at least one substance corresponding to one or more of the at least one material is transferred from or to the location of the second (i.e., from the second cell location or to the second cell location). For example, in conjunction with block 1160 of method 1100, a quantity of a substance, approximately equal to the determined move quantity or quantities of the corresponding materials, may be transferred from the first cell to the second cell or vice versa. The method stops at block 1210. Referring now to FIG. 13, an illustrative method 1300 is provided for simulating and/or controlling a nuclear reactor. The method 1300 starts at a block 1305. As illustrated by point A, method 1300 may be preceded by method 1100. Illustrative method 1300 provides an exemplary method of moving matter in a four-cell loop through the reactor. At each cell in the loop, the amount and type of matter moved need not be identical. A person of skill in the art would understand that the four-cell loop may be expanded or contracted as suitable (i.e., include fewer or more cells). At block 1305, approximately the move quantity(-ies) of material(s) is/are transferred to a second cell. For example, in conjunction with block 1160 of method 1100, a quantity of a substance, approximately equal to the determined move quantity or quantities of the corresponding materials, may be transferred from the first cell to the second cell. At block 1310, second move quantity(-ies) is (are) determined. The second move quantity or quantities may be calculated in any way as described above. At block 1315, amount(s) in the second cell is (are) adjusted by the second move quantity(-ies). At block 1320, approximately the second move quantity(-ies) of material(s) are transferred from the second cell to a third cell. At block 1325, third move quantity(-ies) is (are) determined. At block 1330, amount(s) in the third cell is (are) adjusted by the third move quantity(-ies). At block 1335, approximately the third move quantity(-ies) of material(s) is (are) transferred from the third cell to a fourth cell. At block 1340, fourth move quantity(-ies) is (are) determined. At block 1345, amount(s) in the fourth cell is (are) adjusted by the fourth move quantity(-ies). At block 1350, approximately the fourth move quantity(-ies) of material(s) is (are) transferred from the fourth cell to a first cell The method stops at block 1350. Referring now to FIG. 14, an illustrative method 1400 is provided for simulating and/or controlling a nuclear reactor. The method 1400 starts at a block 1405. As illustrated by point A, method 1400 may be preceded by method 1100. Illustrative method 1400 illustrates, inter alia, mixing quantities of one or materials from a first cell and a third cell, and transferring at least a portion of the mixture back to the first cell. One or more additional iterations of neutron flux and transmutation calculations may optionally occur. A person of skill in the art would understand that this illustrative method could be expanded or contracted to include various mixing methods using fewer or more cells. At block 1405, approximately the move quantity(-ies) of materials is (are) transferred to a second cell. At block 1410, second move quantity(-ies) is (are) determined. At block 1415, amount(s) in the second and third cells is (are) adjusted by the second move quantity(-ies). At block 1420, approximately the second move quantity(-ies) are transferred to/from a third cell and from/to the second cell. At block 1425, a new average rate(s) of change of the amount(s) of the material(s) in the first cell based on current amount(s) of the material(s) and an updated flux in the first cell is (are) determined. At block 1430, new updated amount(s) in the first cell for the material(s) is (are) determined based on the new average rate(s) of change. At block 1435, third move quantity (-ies) is (are) determined. At block 1440, amount(s) in the second and first cell is (are) adjusted by the third move quantity(-ies). At block 1445, approximately the third move quantity (-ies) is (are) transferred from the second cell to the first cell. The method stops at block 1445. Referring now to FIG. 15, an illustrative method 1500 is provided for simulating and/or controlling a nuclear reactor. As illustrated by point A, method 1500 may be preceded by method 1100. Illustrative method 1500 provides an exemplary method of, inter alia, transferring quantities of one or more materials to a cell location for holding (e.g., a holding tank or reservoir, etc.). While the material is in the holding cell location, one or more additional iterations of neutron flux and transmutation calculations may optionally occur. Material may also be transferred out of the holding cell (e.g., to a location that is not represented by the first cell). A person of skill in the art would understand that this illustrative method could be expanded or contracted to include various holding methods using fewer or more cells. The method 1500 starts at a block 1505. At block 1505, approximately the move quantity(-ies) of material(s) is (are) transferred to a second cell. At block 1510, second move quantity(-ies) is (are) determined. At block 1515, amount(s) in the second cell is (are) adjusted by the second move quantity(-ies). At block 1520, approximately the second move quantity(-ies) of material(s) is (are) transferred from the second cell. At block 1525, new average rate(s) of change of the amount(s) of the material(s) in the first cell is (are) determined based on current amount(s) of the material(s) and an updated flux in the first cell. At block 1530, new updated amount(s) in the first cell for the material(s) is (are) determined based on the new average rate(s) of change. At block 1535, an average rate(s) of change of the amount(s) of the material(s) in the second cell is (are) determined based on current amount(s) of the material(s) and a flux in the second cell. At block 1540, updated amount(s) in the second cell for the material(s) is (are) determined based on the average rate of change in the second cell. At block 1545, third move quantity(-ies) is (are) determined. At block 1550, amount(s) in the first and second cells is (are) adjusted by the third move quantity(-ies). At block 1555, approximately the third move quantity(-ies) of material(s) is (are) transferred from the second cell to the first cell. The method stops at block 1555. Referring now to FIG. 16, an illustrative method 1600 is provided for simulating and/or controlling a nuclear reactor. Illustrative method 1600 differs from illustrative method 1110, but some steps may be similar. For example, illustrative method 1600 provides an example of a continuous move of material(s) rather than discrete moves. The method 1600 starts at a block 1605. At block 1605, a flux in a first cell is determined based on amount(s) of at least one material in the first cell. As discussed above, the flux determination may be further based on the amounts of more than one material in the first cell. Also, the flux determination may be further based on the amounts one or more materials in one or more other cells. This block is similar to block 1105. At block 1610, average rate(s) of change of the amount(s) of the material(s) in the first cell is (are) determined based on previous amount(s) of the material(s) and a flux in the first cell. This block is similar to block 1110. At block 1615, at least one move rate for the material(s) in the first cell is (are) determined. A move rate may be any rate of movement of one or more materials such as a quantity of a material that is desired to be moved into or out of a cell. In this block one or more move rates may each apply to one or more materials in the first cell. A move rate may be determined in response to one or more reactor parameters such as a flux or a fluence, a power level (local or overall), a temperature, etc. A reactor parameter may be compared to a threshold or set point for that parameter. This block may be repeated as suitable, e.g., for each of one or more materials in the first cell. At block 1620, the average rate(s) of change in the first cell is (are) adjusted based on the move rate(s) for the material(s) in the first cell. For example, an average rate of change of a material in the first cell may be adjusted (increased or decreased) based on a determined move rate. The adjustment may be made to a single average rate of change or to individual rates of change which are averaged (e.g., in a straight average or a weighted average). The average rates of change for more than one material may be solved individually or simultaneously (such as when coupled through transmutation equations). At block 1625, updated amount(s) in the first cell is (are) determined based on the adjusted average rate(s) of change. In an embodiment, a material may be moved outside the reactor. In this case, blocks 1630 through 1645 may be skipped. At block 1630, a flux in a second cell is determined based on amount(s) of at least one material in the second cell. As discussed above, the flux determination may be further based on the amounts of more than one material in the first cell. Also, the flux determination may be further based on the amounts one or more materials in one or more other cells. At block 1635, average rate(s) of change of the amount(s) of the material(s) in the second cell is (are) determined based on previous amount(s) of the material(s) and a flux in the second cell. At block 1640, the average rate(s) of change in the second cell is (are) determined based on move rate(s) of at least one material. At block 1645, updated amount(s) in the second cell is (are) based on the adjusted average rate(s) of change. At block 1650, a control action for a nuclear reactor is determined. In an embodiment, this block is optional. At block 1655, a control action for nuclear reactor is performed. In an embodiment, this block is optional. At block 1660, material(s) is (are) transferred to/from the first cell at a transfer rate approximately equivalent to the move rate(s) of the material(s). As with block 1160, an actual amount of at least one substance is transferred, but the transfer is at a transfer rate equivalent to the appropriate move rate. In an embodiment such as a simulation or evaluation of a reactor design, this step is optional. The method stops at block 1660, but may continue to point B as indicated in other methods in other figures. Referring now to FIG. 17, an illustrative method 1700 is provided for simulating and/or controlling a nuclear reactor. The method 1700 starts at a block 1705. As illustrated by point B, method 1700 may be preceded by method 1600. At block 1705, material(s) is (are) transferred from/to a second cell at the transfer rate(s) approximately equivalent to the move rate(s) of the materials. In this block, an actual of amount of at least one substance corresponding to one or more of the at least one material is transferred from or to the location of the second (i.e., from the second cell location or to the second cell location) at the move rate(s). For example, in conjunction with block 1660 of method 1600, a quantity of a substance, approximately equal to the determined move quantity or quantities of the corresponding materials, may be transferred from the first cell to the second cell or vice versa at the appropriate move rate(s). The method stops at block 1705. Referring now to FIG. 18, an illustrative method 1800 is provided for simulating and/or controlling a nuclear reactor. The method 1800 starts at a block 1805. As illustrated by point B, method 1800 may be preceded by method 1600. Illustrative method 1800 provides an exemplary method of moving matter at various rates in a four-cell loop through the reactor. At each cell in the loop, the rate and type of matter moved need not be identical. A person of skill in the art would understand that the four-cell loop may be expanded or contracted as suitable (i.e., include fewer or more cells). At block 1805, material(s) is (are) transferred to the second cell at transfer rate(s) approximately equivalent to the move rate(s) of the materials. At block 1810, a second (set of) move rate(s) for material(s) in the second cell is (are) determined. At block 1815, further adjustment(s) to average rate(s) of change in the second cell is (are) made based on the second (set of) move rate(s) of the material(s). At block 1820, material(s) is (are) transferred from the second cell to a third cell at approximately the second (set of) move rate(s). At block 1825, move rate(s) for material(s) in the third cell is (are) determined. At block 1830, average rate(s) of change for material(s) in the third cell is (are) adjusted by the second (set of) move rates for the second cell and/or the determined move rates of the third cell. At block 1835, material(s) is (are) transferred from the third cell to a fourth cell at approximately the determined move rate(s) of the third cell. At block 1840, move rate(s) for material(s) in the fourth cell is (are) determined. At block 1845, average rate(s) of change for material(s) in the fourth cell is (are) adjusted by the determined move rate(s) of the third cell and/or the determined move rates of the fourth cell. At block 1850, material(s) is (are) transferred from the fourth cell to the first cell at approximately the determined move rates of the fourth cell. The method stops at block 1850. Referring now to FIG. 19, an illustrative method 1900 is provided for simulating and/or controlling a nuclear reactor. Illustrative method 1900 illustrates, inter alia, mixing quantities of one or materials from a first cell and a third cell, and transferring at least a portion of the mixture back to the first cell. The transfers occur at various rates. One or more additional iterations of neutron flux and transmutation calculations may optionally occur during the transfer. A person of skill in the art would understand that this illustrative method could be expanded or contracted to include various mixing methods using fewer or more cells. The method 1900 starts at a block 1905. As illustrated by point B, method 1900 may be preceded by method 1600. At block 1905, material(s) is (are) transferred to the second cell at transfer rate(s) approximately equivalent to the move rate(s) of the materials. At block 1910, a second (set of) move rate(s) for material(s) of the second cell is (are) determined. At block 1915, further adjustments are made to average rate(s) of change in the second cell and to the average rate(s) of change in a third cell based on the second (set of) move rate(s) of the material(s). At block 1920, material(s) is (are) transferred at approximately the second (set of) move rate(s) to/from a third cell from/to the second cell. At block 1925, new average rate(s) of change of the amount(s) of the material(s) in the first cell is (are) determined based on current amount(s) of the material(s) and an updated flux in the first cell. At block 1930, new updated amount(s) in the first cell for the material(s) is (are) determined based on the new average rate(s) of change. At block 1935, a third (set of) move rate(s) is determined for material(s) of the second cell. At block 1940, average rate(s) of change of material(s) in the first and second cells is (are) adjusted using the third (set of) move rate(s). At block 1945, material(s) is (are) are transferred at approximately the third (set of) move rate(s) from the second cell to the first cell. The method stops at block 1945. Referring now to FIG. 20, an illustrative method 2000 is provided for simulating and/or controlling a nuclear reactor. Illustrative method 2000 provides an exemplary method of, inter alia, transferring quantities of one or more materials to a cell location for holding (e.g., a holding tank or reservoir, etc.) at various rates. While the material is being transferred to/from the holding cell location, one or more additional iterations of neutron flux and transmutation calculations may optionally occur. Material may also be transferred out of the holding cell (e.g., to a location that is not represented by the first cell). The method 2000 starts at a block 2005. As illustrated by point B, method 2000 may be preceded by method 1600. At block 2005, material(s) is (are) transferred to the second cell at transfer rate(s) approximately equivalent to the move rate(s) of the materials. At block 2010, a second (set of) move rate(s) for material(s) of the second cell is determined. At block 2015, further adjustment(s) is (are) made to average rate(s) of change in the second cell based on the second (set of) move rate(s) of the material(s). At block 2020, material(s) is (are) transferred at approximately the second (set of) move rate(s) from the second cell. At block 2025, new average rate(s) of change of the amount(s) of the material(s) in the first cell is (are) determined based on current amount(s) of the material(s) and an updated flux in the first cell. At block 2030, new updated amount(s) in the first cell is (are) determined for the material(s) based on the new average rate(s) of change. At block 2035, new average rate(s) of change of the amount(s) of the material(s) in the second cell is (are) determined based on current amount(s) of the material(s) and a flux in the second cell. At block 2040, a third (set of) move rate(s) for materials in the second cell is determined. At block 2045, the average rate(s) of change of the amount(s) of material(s) in the first cell is (are) determined based on the third (set of) move rate(s) for materials in the second cell. At block 2050, the average rate(s) of change of the amount(s) of material(s) in the second cell is (are) determined based on the third (set of) move rate(s) for materials in the second cell. At block 2055, material(s) is (are) transferred at approximately the third (set of) move rate(s) from the second cell to the first cell. The method stops at block 2055. Referring now to FIG. 21, an illustrative method 2100 is provided for simulating and/or controlling a nuclear reactor. As discussed elsewhere herein, neutron transport equations may use cross sectional data for some or all of the materials in the reactor. Method 2100 illustrates a non-limiting example method that may have one or more of the following benefit. It may reduce the computational burden, reduce the need for exhaustive cross sectional data for each reaction for each target particle across a spectrum of incident particle energies, and/or improve accuracy of current methods. The method 2100 starts at a block 2105. At block 2105, a neighbor to a principal material in a first cell is selected. A principal material is a material of one or more materials in a reactor or reactor cell that may be represented by an agent material. In an embodiment, principal materials may be fission products (elements, isotopes, and/or isomers of isotopes). A principal material might not be well-characterized with respect to a microscopic property such as microscopic cross sectional data. For example, some of the cross sections for scattering, radiative capture, fission, etc. reactions with neutrons of various energies may not be known. Also, the principal's information may be well-known, but for other reasons (such as reducing computational burden), the principal material may be represented by a neighbor material which will act as an agent. The neighbor may be selected from a plurality of agent materials. In an embodiment, multiple neighbors may be selected from the plurality of agent materials to represent the principal material as agents for more than one property. Agent materials are typically well-characterized with regard to pertinent microscopic cross sectional data. In an embodiment, agent materials are actual materials (e.g., isotopes) with empirically determined microscopic quantities. In a further embodiment, agent materials include one or more fictional materials. A fictional material is essentially a collection of fictional values for various cross sections and optionally other properties. A neighbor may be chosen to act as the agent for the principal on one or more of many criteria. Typically, a neighbor has microscopic properties such that a certain density of the neighbor will have similar macroscopic properties as the existing density of the principal. Thus, a neighbor may be selected to act as an agent for the principal based on a comparison of microscopic properties of the principal to the microscopic properties of each of the neighbors. Microscopic properties may be approximated having one or more discrete values with respect to incident particle (e.g., neutron) energy, or may be evaluated as a function of incident particle energy. In an embodiment, the selection of a neighbor or neighbors is limited to a selection from one or more agent materials that are also fission products of the fissions of one or more fissile materials. The fission products may further be limited to fissions induced by neutrons and/or neutrons of certain energy levels. In an embodiment further limiting the selection, potential neighbors may be chosen from agent materials under the same “hump” as the principal material of a fission yield curve (e.g., left curve portion 912 or right curve portion 914 of fission yield curve 900 illustrated in FIG. 9). In an embodiment, the number of potential principal materials is larger than the number of agent materials. For example, the known fission product isotopes number in the thousands. In a further embodiment, the number of agent materials is limited to a relatively small number (e.g., under 100, 50, 30, or 20). In an embodiment, the number of agent materials is limited to 12. In an embodiment, this block is performed by exemplary method 2200 described below. At block 2110, a proxy amount of the selected neighbor or neighbors is determined. As discussed above, a neighbor might have microscopic properties such that a certain density of the neighbor will have similar macroscopic properties as the existing density of the principal. The proxy amount is the amount or density (e.g., concentration) of the neighbor that will serve to represent the principal in a given concentration. As with microscopic properties, macroscopic properties may be approximated as one or more discrete values or as a function of incident particle energy. At block 2115, blocks 2105 and 2110 are repeated for each of a plurality of principal materials in the first cell. In this block, a neighbor or neighbors is/are selected to act as an agent(s) for each of the plurality of principal materials (which may or may not make up all of the materials in the cell). A given agent material may be selected as a neighbor for more than one principal material. Other agent materials may not be selected to be any principal's neighbor. Proxy amounts of each agent are determined for each principal to which the agent is a neighbor. At block 2120, a summed proxy amount for each agent material is determined. In this block, a total proxy amount of each agent is determined based on the proxy amounts for each neighbor of the agent material. For example, suppose an agent material was selected to be the neighbor of three different principal materials. After performance of the previous blocks, the agent may have three proxy amounts (one for each principal). In this block, a summed proxy amount is determined based on the three proxy amounts (e.g., by summing them). At block 2125, a flux in the first cell is determined based on the summed proxy amounts of each agent material the first cell. As described elsewhere herein, for example, a flux may be determined by a transport calculation (e.g., solving neutron transport equations to determine a neutron flux) and may be further based on the summed proxy amounts of each agent material in one or more other cells. The flux may be approximated by one or more discrete values, or may be a continuous function, thus describing a flux spectrum. The flux may be space and/or energy dependent. The flux may be determined by numerical analysis methods including Monte Carlo methods. The average rate of change of the flux may be a weighted average (e.g., as determined by a Runge Kutta method or any other method). The flux may be dependent upon the amount of each of one or more materials in the first cell. Instead of using the actual amounts (e.g., concentrations) of each material in the cell, the calculation instead uses the summed proxy amounts of each agent material. Thus, the cross sectional data and concentrations of the principal materials are accounted for in a flux determination (e.g., neutron transport calculations) by agents having proxy concentrations. In embodiments where the number of agents is relatively small, the computational burden may be reduced significantly. At block 2130, an updated amount of one or more materials (principal or agent) is determined based on the previous amount of the materials and the flux (e.g., the estimated average flux) in the first cell. The one or more materials may be a subset of the materials in the cell. For example, updated amounts may be determined by a transmutation calculation, which may take into account production rates (e.g., based on reactions rates such as fission rates) and decay rates (e.g., using decay constants). The updated amounts for the one or more materials may be solved individually or simultaneously (such as when coupled through transmutation equations). The calculation may include calculating the updated amount based on a specified length of time. At block 2135, a control action for a nuclear reactor is determined. As described above, the control action may be a change (positive or negative) to a local neutronic reactivity of a reactor using any neutron affecting or absorbing features such as movement of neutron absorbing materials or fluids, control rods, etc.; a change in one or more various flow rates for any reason including but not limited to localized or overall reactor power, e.g., the flow of heat absorbing material (e.g., coolant) through the reactor or portions of the reactor by ordering changes in reactor coolant pump operation and/or various valve positions in the reactor system, including but not limited to reactor closures or reactor coolant shutoff valves, steam shutoff valves, etc.; a change in one or more breaker positions (e.g., reactor coolant pump power supply breakers, steam turbine-generator output breakers, etc.); or the like. Other control actions will be apparent to persons skilled in the art based on the teachings herein. The determined control action may be displayed to a user. At block 2140, a control action for the nuclear reactor is performed. As described above, this performance may be automatic or manual. The method stops at block 2140. Referring now to FIG. 22, an illustrative method 2200 is provided for simulating and/or controlling a nuclear reactor. In an embodiment, method 2200 is used to perform block 2105 above. The method 2200 starts at a block 2205. At block 2205, a plurality of potential neighbors is identified. Continuing the description of block 2205, potential neighbors may be limited to materials that are fission products of certain isotopes, perhaps induced by incident particles of a certain energy. In an embodiment, potential neighbors for a given principal may be limited to fission products under the same “hump” of a fission yield curve or curves as the principal. The fission yield curve of interest might be, for example, the curve of one particular fissile material's fission reaction or the curves of multiple fission reactions in any combination of incident particle energy and fissile material. Potential neighbors may also be limited to materials which are characterized to the extent necessary to be suitable as agent materials. In an embodiment, the plurality of potential neighbors for a given principal is chosen by identifying some number (e.g., three) of agent materials having atomic mass numbers (A) “most similar” to that of the principal material. The “most similar” decision may be restricted to agent materials having larger (or smaller) atomic mass numbers. Also, the “most similar” decision may be forced to take at least one smaller and one larger (in atomic mass number) agent material. Potential neighbors having a microscopic cross section of zero or close to zero may be ruled out in some embodiments. At block 2210, a neighbor is selected from the plurality of potential neighbors. Once a plurality of potential neighbors is identified, one or more neighbors may be selected from the plurality. In an embodiment, neighbors may be selected by comparing one or more microscopic properties such as a cross section. Thus, the number of comparisons needed to select a neighbor is limited by the number of materials determined to be potential neighbors in block 2205. The method stops at block 2210. In an embodiment, potential neighbors may be identified by comparing one or more microscopic properties such as a cross section. Potential neighbors having a microscopic cross section of zero or close to zero may be ruled out in some embodiments. With respect to the use of substantially any plural and/or singular terms herein, those having skill in the art can translate from the plural to the singular and/or from the singular to the plural as is appropriate to the context and/or application. The various singular/plural permutations are not expressly set forth herein for sake of clarity. While particular aspects of the present subject matter described herein have been shown and described, it will be apparent to those skilled in the art that, based upon the teachings herein, changes and modifications may be made without departing from the subject matter described herein and its broader aspects and, therefore, the appended claims are to encompass within their scope all such changes and modifications as are within the true spirit and scope of the subject matter described herein. Furthermore, it is to be understood that the invention is defined by the appended claims. It will be understood by those within the art that, in general, terms used herein, and especially in the appended claims (e.g., bodies of the appended claims) are generally intended as “open” terms (e.g., the term “including” should be interpreted as “including but not limited to,” the term “having” should be interpreted as “having at least,” the term “includes” should be interpreted as “includes but is not limited to,” etc.). It will be further understood by those within the art that if a specific number of an introduced claim recitation is intended, such an intent will be explicitly recited in the claim, and in the absence of such recitation no such intent is present. For example, as an aid to understanding, the following appended claims may contain usage of the introductory phrases “at least one” and “one or more” to introduce claim recitations. However, the use of such phrases should not be construed to imply that the introduction of a claim recitation by the indefinite articles “a” or “an” limits any particular claim containing such introduced claim recitation to inventions containing only one such recitation, even when the same claim includes the introductory phrases “one or more” or “at least one” and indefinite articles such as “a” or “an” (e.g., “a” and/or “an” should typically be interpreted to mean “at least one” or “one or more”); the same holds true for the use of definite articles used to introduce claim recitations. In addition, even if a specific number of an introduced claim recitation is explicitly recited, those skilled in the art will recognize that such recitation should typically be interpreted to mean at least the recited number (e.g., the bare recitation of “two recitations,” without other modifiers, typically means at least two recitations, or two or more recitations). Furthermore, in those instances where a convention analogous to “at least one of A, B, and C, etc.” is used, in general such a construction is intended in the sense one having skill in the art would understand the convention (e.g., “a system having at least one of A, B, and C” would include but not be limited to systems that have A alone, B alone, C alone, A and B together, A and C together, B and C together, and/or A, B, and C together, etc.). In those instances where a convention analogous to “at least one of A, B, or C, etc.” is used, in general such a construction is intended in the sense one having skill in the art would understand the convention (e.g., “a system having at least one of A, B, or C” would include but not be limited to systems that have A alone, B alone, C alone, A and B together, A and C together, B and C together, and/or A, B, and C together, etc.). It will be further understood by those within the art that virtually any disjunctive word and/or phrase presenting two or more alternative terms, whether in the description, claims, or drawings, should be understood to contemplate the possibilities of including one of the terms, either of the terms, or both terms. For example, the phrase “A or B” will be understood to include the possibilities of “A” or “B” or “A and B.” With respect to the appended claims, those skilled in the art will appreciate that recited operations therein may generally be performed in any order. Examples of such alternate orderings may include overlapping, interleaved, interrupted, reordered, incremental, preparatory, supplemental, simultaneous, reverse, or other variant orderings, unless context dictates otherwise. With respect to context, even terms like “responsive to,” “related to,” or other past-tense adjectives are generally not intended to exclude such variants, unless context dictates otherwise. The herein described subject matter sometimes illustrates different components contained within, or connected with, different other components. It is to be understood that such depicted architectures are merely exemplary, and that in fact many other architectures may be implemented which achieve the same functionality. In a conceptual sense, any arrangement of components to achieve the same functionality is effectively “associated” such that the desired functionality is achieved. Hence, any two components herein combined to achieve a particular functionality can be seen as “associated with” each other such that the desired functionality is achieved, irrespective of architectures or intermedial components. Likewise, any two components so associated can also be viewed as being “operably connected”, “operatively coupled,” or “operably coupled,” to each other to achieve the desired functionality, and any two components capable of being so associated can also be viewed as being “operably couplable,” to each other to achieve the desired functionality. Specific examples of operably couplable include but are not limited to physically mateable and/or physically interacting components, and/or wirelessly interactable, and/or wirelessly interacting components, and/or logically interacting, and/or logically interactable components. While various aspects and embodiments have been disclosed herein, other aspects and embodiments will be apparent to those skilled in the art. The various aspects and embodiments disclosed herein are for purposes of illustration and are not intended to be limiting, with the true scope and spirit being indicated by the following claims. |
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claims | 1. A method of decontaminating a metal surface, wherein the metal surface is located on a component within a nuclear plant and is covered with a metal oxide layer containing radioactive substances, and wherein the method comprises the following steps:a) an oxidation step in which the metal oxide layer is contacted with an aqueous oxidation solution comprising an oxidant;b) a decontamination step wherein the metal oxide layer treated in the oxidation step is contacted with an aqueous solution of an organic acid to dissolve the metal oxide layer, thereby forming a decontamination solution that contains the organic acid as well as metal ions and the radioactive substances, and wherein the decontamination solution is passed over an ion exchanger to immobilize the metal ions and the radioactive substances;wherein in step b) an oxidant selected from the group consisting of oxygen, air, hydrogen peroxide and ozone is dosed into the decontamination solution;wherein the decontamination solution contains both nickel ions and iron ions and the nickel ions, in the decontamination step, are removed earlier from the decontamination solution than the iron ions; andwherein the decontamination solution is passed over a cation exchanger material, binding the nickel ions on the cation exchanger material, while the iron ions remain in the decontamination solution in the form of an anionic Fe(III) complex. 2. The method according to claim 1, characterized in that, in the decontamination step, only a plant-internal ion exchanger is used. 3. The method according to claim 1, characterized in that oxidant dosing is controlled and monitored exclusively with facilities of the nuclear plant. 4. The method according to claim 1, characterized in that the method is used for large-scale system decontamination. 5. The method according to claim 1, characterized in that the component is located in a coolant system of a nuclear power plant and comprises one or more of the following components: a pipeline for a primary coolant, a reactor pressure vessel, a coolant pump, a steam generator, a pressure reduction facility and/or at least one auxiliary system connected to the coolant system. 6. The method according to claim 1, characterized in that, during the decontamination step, a purification rate is in a range of between 0.1 h-1 and 0.7 h-1, the purification rate being defined as a ratio of a flow capacity of the ion exchanger to volume of a system to be decontaminated. 7. The method according to claim 1, characterized in that dosing of the oxidant into the decontamination solution is controlled such that a dissolution rate of the metal oxide layer of 400 ppm Fe/h is not exceeded. 8. The method according to claim 1, characterized in that dosing of the oxidant into the decontamination solution is controlled such that a concentration of Fe(II) ions in the decontamination solution of 50 ppm is not exceeded. 9. The method according to claim 1, characterized in that dosing of the oxidant into the decontamination solution is controlled dependent on available flow capacity of the ion exchanger, system volume and area of the metal surface to be decontaminated. 10. The method according to claim 1, characterized in that dosing of the oxidant into the decontamination solution is controlled such that a ratio of a dissolution rate of the metal oxide layer to a purification rate is at most 25 ppm Fe, the purification rate being defined as the ratio of a flow capacity (m3/h) of the ion exchanger to a volume (m3) of a system to be decontaminated. 11. The method according to claim 1, characterized in that hydrogen peroxide is used as an oxidant and dosed into the decontamination solution in an amount of between 0.1 and 1 L per h per 100 m3 of system volume. 12. The method according to claim 1, characterized in that air, oxygen or ozone is used as an oxidant and that the oxidant is dosed by forced ventilation of a proportion of the decontamination solution. 13. The method according to claim 1, characterized in that air, oxygen or ozone is directly introduced into the decontamination solution under pressure. 14. The method according to claim 1, characterized in that the oxidant is continuously dosed into the decontamination solution. 15. The method according to claim 1, characterized in that the oxidant is dosed into the decontamination solution in a time-delayed manner after a beginning of the decontamination step. 16. The method according to claim 1, characterized in that dosing of the oxidant into the decontamination solution is controlled by monitoring one or more physical and/or electrochemical parameters of the decontamination solution. 17. The method according to claim 16, characterized in that a reduction-oxidation potential, a corrosion potential, a conductivity and/or a total oxygen content of the decontamination solution are monitored and used to control the dosing of the oxidant. 18. The method according to claim 17, characterized in that the reduction-oxidation potential and/or the corrosion potential are set such that the organic acid cannot cause a base metal attack on the metal surface. 19. The method according to claim 17, characterized in that oxidant dosing is controlled such that a limit value for the corrosion potential and/or the reduction-oxidation potential of the decontamination solution of 0 mV is not undershot. 20. The method according to claim 17, characterized in that dosing of the oxidant into the decontamination solution begins as soon as the corrosion potential and/or the reduction-oxidation potential of the decontamination solution has dropped to a value in a range from 20 to 180 mV. 21. The method according to claim 1, characterized in that the decontamination step is followed by a purification step in which the organic acid contained in the decontamination solution is removed and optionally decomposed, with the oxidation step, the decontamination step and the purification step being optionally repeated. 22. The method according to claim 1, wherein the metal surface is located on a component within a coolant system of a nuclear power plant. 23. The method according to claim 4, characterized in that the method is used for full system decontamination. 24. The method according to claim 22, characterized in that, during the decontamination step, a purification rate is in a range of between 0.1 h-1 and 0.7 h-1, the purification rate being defined as a ratio of a flow capacity of the ion exchanger to volume of the coolant system to be decontaminated. 25. The method according to claim 22, characterized in that air, oxygen or ozone is used as an oxidant and that the oxidant is dosed by forced ventilation of a proportion of the decontamination solution diverted from the coolant system. 26. The method according to claim 15, characterized in that the dosing of the oxidant begins after the decontamination solution has reached a total iron concentration of at least 20 ppm. |
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claims | 1. A beam processing system comprising a disk mounted thereon with a plurality of processing objects on the same circumference, a rotation drive mechanism for rotating said disk about a disk axis, a reciprocating drive mechanism for causing said disk, while rotating, to perform a reciprocating scan motion in a direction perpendicular to said disk axis within a stroke range defined by an inner overscan position and an outer overscan position, and a controller for controlling at least said reciprocating drive mechanism, said beam processing system causing said plurality of processing objects to pass through an irradiation position of a processing beam by rotation and the reciprocating scan motion of said disk, thereby irradiating the processing beam onto said plurality of processing objects,wherein said beam processing system further comprises a beam width measuring unit for measuring a beam width of said processing beam,said controller sets said inner overscan position and said outer overscan position depending on a measured value of said beam width or a predetermined value of said beam width, andsaid controller, based on the number of rotation of said disk per unit time, a scan speed and the number of reciprocating scan times of said reciprocating scan motion, a reversal start timing of said disk at at least one of said inner overscan position and said outer overscan position, and said measured value of the beam width or said predetermined value of the beam width, controls said reciprocating drive mechanism so as to ensure an overlap region between a last and a current processing beam irradiation region on each of said plurality of processing objects, said overlap region overlapping at least half of said last processing beam irradiation region. 2. The beam processing system according to claim 1, wherein said controller sets said inner overscan position and said outer overscan position depending on said measured value of the beam width and measured values of one end position and the other end position in a scan direction on a cross section of the processing beam. 3. The beam processing system according to claim 1, wherein said controller determines the scan speed and the number of reciprocating scan times of said reciprocating scan motion depending on a required total beam irradiation amount. 4. The beam processing system according to claim 1, wherein said controller ensures said overlap region between the last and current processing beam irradiation regions on each of said plurality of processing objects by regularly delaying per scan the reversal start timing of said disk at said at least one of said inner overscan position and said outer overscan position. 5. The beam processing system according to claim 4, further comprising a target detecting unit provided at a position adjacent to said disk for detecting an initial position detection target portion provided at a predetermined position of said disk while said disk is rotating, and outputting a detection signal,wherein said controller delays said reversal start timing based on said detection signal. 6. The beam processing system according to claim 1, wherein said controller ensures said overlap region between the last and current processing beam irradiation regions on each of said plurality of processing objects by delaying per scan the reversal start timing of said disk at said at least one of said inner overscan position and said outer overscan position by a delay time (T/Nx) derived by dividing a rotation period T of said disk by a value Nx set based on the number of reciprocating scan times N. 7. The beam processing system according to claim 6, further comprising a target detecting unit provided at a position adjacent to said disk for detecting an initial position detection target portion provided at a predetermined position of said disk while said disk is rotating, and outputting a detection signal,wherein said controller delays said reversal start timing based on said detection signal. 8. The beam processing system according to claim 6, wherein when said number of reciprocating scan times N is large, said controller sets said value Nx so as to increase by stages as the number of reciprocating times approaches the number N. 9. The beam processing system according to claim 1, wherein said controller ensures said overlap region between the last and current processing beam irradiation regions on each of said plurality of processing objects by randomly controlling, based on random numbers, the reversal start timing of said disk at said at least one of said inner overscan position and said outer overscan position. 10. The beam processing system according to claim 1, wherein said controller ensures said overlap region between the last and current processing beam irradiation regions on each of said plurality of processing objects by controlling the reversal start timing of said disk at said at least one of said inner overscan position and said outer overscan position, based on a programmed relationship of said number of rotation of said disk, said number of reciprocating scan times, said scan speed, and said measured value of the beam width. 11. The beam processing system according to claim 1, wherein said controller stores a relationship between said number of rotation of said disk and said number of reciprocating scan times as table data in a storage device in advance and ensures said overlap region between the last and current processing beam irradiation regions on each of said plurality of processing objects by controlling, based on said table data, the reversal start timing of said disk at said at least one of said inner overscan position and said outer overscan position. 12. A beam processing method that causes a disk mounted thereon with a plurality of processing objects on the same circumference to rotate about a disk axis, causes said disk, while rotating, to perform a reciprocating scan motion in a direction perpendicular to said disk axis within a stroke range defined by an inner overscan position and an outer overscan position, and causes said plurality of processing objects to pass through an irradiation position of a processing beam by rotation and the reciprocating scan motion of said disk, thereby irradiating the processing beam onto said plurality of processing objects, said beam processing method comprising:measuring a beam width of said processing beam;setting said inner overscan position and said outer overscan position depending on a measured value of said beam width or a predetermined value of said beam width; andbased on the number of rotation of said disk per unit time, a scan speed and the number of reciprocating scan times of said reciprocating scan motion, a reversal start timing of said disk at at least one of said inner overscan position and said outer overscan position, and said measured value of the beam width or said predetermined value of the beam width, controlling said reciprocating scan motion so as to ensure an overlap region between a last and a current processing beam irradiation region on each of said plurality of processing objects, said overlap region overlapping at least half of said last processing beam irradiation region. 13. The beam processing method according to claim 12, comprising setting said inner overscan position and said outer overscan position depending on said measured value of the beam width and measured values of one end position and the other end position in a scan direction on a cross section of the processing beam. 14. The beam processing method according to claim 12, comprising determining the scan speed and the number of reciprocating scan times of said reciprocating scan motion depending on a required total beam irradiation amount. 15. The beam processing method according to claim 12, comprising ensuring said overlap region between the last and current processing beam irradiation regions on each of said plurality of processing objects by regularly delaying per scan the reversal start timing of said disk at said at least one of said inner overscan position and said outer overscan position. 16. The beam processing method according to claim 15, comprising using, as a reference for delaying said reversal start timing, a detection signal obtained by detecting an initial position detection target portion provided at a predetermined position of said disk while said disk is rotating. 17. The beam processing method according to claim 12, comprising ensuring said overlap region between the last and current processing beam irradiation regions on each of said plurality of processing objects by delaying per scan the reversal start timing of said disk at said at least one of said inner overscan position and said outer overscan position by a delay time (T/Nx) derived by dividing a rotation period T of said disk by a value Nx set based on the number of reciprocating scan times N. 18. The beam processing method according to claim 17, comprising using, as a reference for delaying said reversal start timing, a detection signal obtained by detecting an initial position detection target portion provided at a predetermined position of said disk while said disk is rotating. 19. The beam processing method according to claim 17, comprising, when said number of reciprocating scan times N is large, setting said value Nx so as to increase by stages as the number of reciprocating times approaches the number N. 20. The beam processing method according to claim 12, comprising ensuring said overlap region between the last and current processing beam irradiation regions on each of said plurality of processing objects by randomly controlling, based on random numbers, the reversal start timing of said disk at said at least one of said inner overscan position and said outer overscan position. 21. The beam processing method according to claim 12, comprising ensuring said overlap region between the last and current processing beam irradiation regions on each of said plurality of processing objects by controlling the reversal start timing of said disk at said at least one of said inner overscan position and said outer overscan position, based on a programmed relationship of said number of rotation of said disk, said number of reciprocating scan times, said scan speed, and said measured value of the beam width. 22. The beam processing method according to claim 12, comprising preparing a relationship between said number of rotation of said disk and said number of reciprocating scan times as table data in advance and ensuring said overlap region between the last and current processing beam irradiation regions on each of said plurality of processing objects by controlling, based on said table data, the reversal start timing of said disk at said at least one of said inner overscan position and said outer overscan position. 23. A beam processing system comprising a disk mounted thereon with a plurality of processing objects on the same circumference, a rotation drive mechanism for rotating said disk about a disk axis, a reciprocating drive mechanism for causing said disk, while rotating, to perform a reciprocating scan motion in a direction perpendicular to said disk axis within a stroke range defined by an inner overscan position and an outer overscan position, and a controller for controlling at least said reciprocating drive mechanism, said beam processing system causing said plurality of processing objects to pass through an irradiation position of a processing beam by rotation and the reciprocating scan motion of said disk, thereby irradiating the processing beam onto said plurality of processing objects,wherein said beam processing system further comprises a beam width measuring unit for measuring a beam width of said processing beam,said controller sets said inner overscan position and said outer overscan position depending on a measured value of said beam width or a predetermined value of said beam width, andsaid controller sets a scan speed of said reciprocating scan motion based on selection and setting of the number of rotation of said disk per unit time so that a distance of said reciprocating scan motion during one rotation of said disk becomes smaller than said measured value of the beam width or said predetermined value of the beam width and, after determining the number of reciprocating scan times, sets a reversal start timing of said reciprocating scan motion, thereby controlling said reciprocating drive mechanism so that irradiation of the processing beam is performed onto said plurality of processing objects so as to provide an overlap region overlapping at least half of a last irradiation region of the processing beam on each of said plurality of processing objects. 24. A beam processing method that causes a disk mounted thereon with a plurality of processing objects on the same circumference to rotate about a disk axis, causes said disk, while rotating, to perform a reciprocating scan motion in a direction perpendicular to said disk axis within a stroke range defined by an inner overscan position and an outer overscan position, and causes said plurality of processing objects to pass through an irradiation position of a processing beam by rotation and the reciprocating scan motion of said disk, thereby irradiating the processing beam onto said plurality of processing objects, said beam processing method comprising:measuring a beam width of said processing beam;setting said inner overscan position and said outer overscan position depending on a measured value of said beam width or a predetermined value of said beam width; andsetting a scan speed of said reciprocating scan motion based on selection and setting of the number of rotation of said disk per unit time so that a distance of said reciprocating scan motion during one rotation of said disk becomes smaller than said measured value of the beam width or said predetermined value of the beam width and, after determining the number of reciprocating scan times, setting a reversal start timing of said reciprocating scan motion, thereby controlling said reciprocating scan motion so that irradiation of the processing beam is performed onto said plurality of processing objects so as to provide an overlap region overlapping at least half of a last irradiation region of the processing beam on each of said plurality of processing objects. 25. A beam processing system comprising a disk mounted thereon with a plurality of processing objects on the same circumference, a rotation drive mechanism for rotating said disk about a disk axis, a reciprocating drive mechanism for causing said disk, while rotating, to perform a reciprocating scan motion in a direction perpendicular to said disk axis within a stroke range defined by an inner overscan position and an outer overscan position, and a controller for controlling at least said reciprocating drive mechanism, said beam processing system causing said plurality of processing objects to pass through an irradiation position of a processing beam by rotation and the reciprocating scan motion of said disk, thereby irradiating the processing beam onto said plurality of processing objects,wherein said beam processing system further comprises a beam width measuring unit for measuring a beam width of said processing beam,said controller sets said inner overscan position and said outer overscan position depending on a measured value of said beam width or a predetermined value of said beam width, andsaid controller sets a scan speed of said reciprocating scan motion based on selection and setting of the number of rotation of said disk per unit time so that a distance of said reciprocating scan motion during one rotation of said disk becomes greater than said measured value of the beam width or said predetermined value of the beam width and, after selecting the number of reciprocating scan times, sets a reversal start timing of said reciprocating scan motion, thereby controlling said reciprocating drive mechanism so that irradiation of the processing beam is performed onto said plurality of processing objects so as to provide an overlap region overlapping at least half of a last irradiation region of the processing beam on each of said plurality of processing objects. 26. A beam processing method that causes a disk mounted thereon with a plurality of processing objects on the same circumference to rotate about a disk axis, causes said disk, while rotating, to perform a reciprocating scan motion in a direction perpendicular to said disk axis within a stroke range defined by an inner overscan position and an outer overscan position, and causes said plurality of processing objects to pass through an irradiation position of a processing beam by rotation and the reciprocating scan motion of said disk, thereby irradiating the processing beam onto said plurality of processing objects, said beam processing method comprising:measuring a beam width of said processing beam;setting said inner overscan position and said outer overscan position depending on a measured value of said beam width or a predetermined value of said beam width; andsetting a scan speed of said reciprocating scan motion based on selection and setting of the number of rotation of said disk per unit time so that a distance of said reciprocating scan motion during one rotation of said disk becomes greater than said measured value of the beam width or said predetermined value of the beam width and, after selecting the number of reciprocating scan times, setting a reversal start timing of said reciprocating scan motion, thereby controlling said reciprocating scan motion so that irradiation of the processing beam is performed onto said plurality of processing objects so as to provide an overlap region overlapping at least half of a last irradiation region of the processing beam on each of said plurality of processing objects. |
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048225550 | description | DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS An embodiment of the present invention will now be described with reference to FIGS. 1 and 2. This is an embodiment of a container for use in placing plate-like objects 1 therein which are produced by cutting used fuel channel boxes as mentioned above. FIG. 1 is a longitudinal section of the embodiment, and FIG. 2 is a horizontal section thereof. Referring to these drawings, reference numeral 5 denotes a container body which consists in this embodiment of a cross-sectionally rectangular box having a height of 412 cm, a width of 227 mm, a length of 300 mm and a wall thickness of 6 mm and made of SUS304 stainless steel. Reference numeral 6 denotes two plate springs each of which consists of a double plate material and extends diagonally in the lengthwise direction of the container 5. The upper end portion of each plate spring is fixed to the wall of the container 5 by welding or by suitable means, such as screws or rivets, and the lower end portion thereof is positioned near an inner surface of the container 5 or contacts the same lightly or urges the same with a suitable level of resilient force. The lower end of each plate spring 6 is preferably in the position which is not higher than the middle of a total height of the container 5. In this embodiment, each plate spring 6 is made of SUS304 stainless steel, and its fixing point is on the portion of the inner wall surface of the container 5 which is vertically about 170 cm away from the lower end thereof, the lower end of the plate spring 6 urging with a force of about 1 kg the portion of the opposite inner wall surface of the container 5 which is vertically about 100 cm away from the lower end thereof. These numerical values are shown as examples, and may be set to suitable different levels as necessary. A container 5 of the above-described construction is set in a storage pool 2 in the same manner as shown in FIG. 5, and a plate-like object 1 produced by cutting a used fuel channel box is brought above the container 5 in the same manner as mentioned above. When the plate-like object 1 is lowered from a position in the upper portion of FIG. 1 into the interior of the container 5, it is pressed by the plate spring 6 as it is guided into the container. The plate-like object 1 is thus placed in the container 5 with the former contacting the inner wall surface of the latter. When a subsequent plate object 1 is then lowered in a suspended state into the container 5 in the same manner, it is placed therein in close contact with the preceding plate-like object through the same actions as described above. The plate-like objects 1 are thus placed successively in tight contact with one another until the last plate-like object 1 is placed in the position shown by a broken line in FIG. 2. In a conventional container of this kind having no plate spring 6 therein, only thirty pieces (corresponding to fifteen channel boxes) at most of plate-like objects 1 can be held, while, in the above embodiment of the present invention, which has the same dimensions as the conventional container, thirty-six (corresponding to eighteen channel boxes) plate-like objects 1 can be held even if the elongated rod referred to in the statement under "Description of the Related Art" is not used. FIG. 3 is a horizontal section of an embodiment of the present invention in which flat plate-like objects, not angle-like objects, are to be placed in tight contact with one another, the construction and operation of this embodiment being similar to those of the first embodiment previously described. In this embodiment, the plate-like objects 1' can be inserted in a tightly laminated state until the last plate-like object 1' is placed in the position shown by a broken line in the same drawing. In the above-described embodiments, two plate springs 6 are arranged in the widthwise direction of a container. One plate spring may also be provided at the widthwise intermediate portion of the interior of the container, or not less than two plate springs at horizontally symmetric portions of the interior thereof. In the above-described embodiments, only one or only one horizontal row of plate springs are provided in the container. The plate springs may also be provided at vertically spaced portions of the interior of the container. The form of the springs may be modified in various manners. The present invention has been described on the basis of its embodiments used as containers for plate-like objects produced by cutting fuel channel boxes taken out of a nuclear reactor. The use of the present invention is not limited to this. It can, of course, be used as a container for general plate-like objects. According to the present invention, plate-like objects can be placed in a container with a high operation efficiency with the former tightly contacting one another and the inner surface of the latter, by merely lowering the plate-like objects successively into the interior of the container. The container according to the present invention is capable of holding a larger number of plate-like objects as compared with a conventional container of this kind of the same dimensions. Therefore, the space utilization rate of the container can be improved. Moreover, the construction of the container is very simple. The present invention is not, of course, limited to the above embodiments; it may be modified in various ways within the scope of the appended claims. |
046860789 | description | DESCRIPTION OF PREFERRED EMBODIMENT Referring first to FIGS. 1 and 2, the reactor vessel shell 10 is shown with its head removed from a circular flange 13. A horizontal outlet nozzel 14 extends through the shell 10 and approximately 120 degrees away from it is a similarly positioned horizontal inlet nozzel 15. Guide studs 16 and 17 are shown screwed in two of many equally spaced stud holes (for example 52 in number) in the flange 13. It will be evident from FIG. 2 that the guide stud 17 and inlet nozzle 15 have been rotated 60 degrees from their true plane so as to be visible in section in FIG. 1 and hence the section line of 1--1 in FIG. 2 is angled. The method and apparatus of the invention are applicable to ultrasonic inspection of the welds within the outlet and inlet nozzles 14 and 15 respectively. The vertical guide studs 16 and 17 extending upwardly from the upper circular vessel flange 13 are submerged in a water pool having a surface 20 many feet above the upper ends of the guide studs 16 and 17. Well above the water pool surface 20 is a refuelling bridge structure suggested schematically at 21 in FIG. 1 from which personnel may work during the assembly and operation of the apparatus of the invention. Referring now to FIG. 3 and to FIGS. 4 to 8 which are sections thereof, a spider assembly designated generally by 23 is shown to include three basically tubular legs 24, 25 and 26 spaced approximately 120 degrees apart. A hub assembly 27 joins the converging ends of the legs 24, 25 and 26 in position such that the outer ends of the legs extend radially outwardly a distance sufficient to span the flange 13 of the reactor vessel. As shown in FIG. 5, the hub assembly includes a pair of bushings 12 coaxial with the center of the spider assembly 23. The tubular leg 24 is shown in FIG. 5 projecting from the hub assembly and is equipped with an air tap 29 and a water tap 30 communicating with its hollow interior. The interior of the leg 24 thus forms a buoyancy chamber. Referring to FIG. 7, a brake shoe 32 operable by a pneumatic cylinder assembly 33 is provided within the hub assembly 27 to operate between the bushings 12. An inner leg mounting block 35 is shown in section in FIG. 6 to illustrate the manner in which each of the legs is affixed to the hub assembly 27 of the spider assembly 23. Closure means 36 shown in FIG. 6 constitutes the end wall of the buoyancy chamber defined by the leg 24. Air taps 38 and 39 on the legs 25 and 26 respectively are similar to the air tap 29 on the leg 24. Associated water taps not visible in the drawings are provided opposite the air taps 38 and 39 for the legs 25 and 26 just as the water tap 30 is located opposite the air tap 29 for the leg 24. Each of the tubular legs 25 and 26 defines its own buoyancy chamber in a manner similar to the leg 24. As shown in FIG. 3, at the outer end of the leg 24 is an articulated clamp 41 operated by a pneumatic cylinder 42. As will be clear from the following description the clamp 41 is adapted to slide on and grip the guide stud 17. As shown in FIG. 4 an end plate of the leg 40 to which the clamp 41 is attached includes a flange spacer 44 which serves as a foot for resting on the reactor flange 13. FIG. 8 illustrates a similar flange spacer 45 affixed to an end plate 46 at the end of the leg 26. A third flange spacer not shown in the drawings is provided on an end plate 47 at the outer end of the third leg 25 of the spider assembly. As shown in FIG. 3 there is mounted to the end plate 47 a clamp 49 operable by a pneumatic air cylinder 50. The clamp 49 is adapted to slide on and grip the guide stud 16 extending upwardly from the flange 13 of the reactor vessel. Each of the air cylinders 33 for the brake shoe 32, and the air cylinders 42 for the clamp 41 and the air cylinder 50 for the clamp 49 is operated by remote control through pneumatic lines which are not shown in the drawings but which extend to an appropriate control station on the refuelling bridge 21 above the surface 20 of the water pool. The bushings 12 of the hub assembly 27 are designed to fit around a vertical column 52 shown in FIG. 9. The total length of the vertical column is approximately six feet. The lower end of the column 52 is attached to a frame 53 of a boom assembly 54 by the use of captive screws 55. The column 52 is provided with a circumferential lower stop 57 which is permanently in place and an upper circumferential stop 58 which is positioned by hose clamps after the vertical column 52 is placed within the bushings 12 of the hub assembly 57. At the extreme upper end of the column 52 above the stop 58 is a socket 59 for receiving one or more positioning rods 60 coaxially attachable to the column remote from the boom assembly 54. The upper end of the positioning rod 60 extends above and adjacent the refuelling bridge structure for appropriate manipulation by the operator. The vertical column 52 is tubular and hollow and defines its own buoyancy chamber provided with air and water taps and appropriate hoses (not shown in the drawings) similar to those provided for the legs 24, 25 and 26 of the spider assembly. The boom assembly 54 includes a boom tube 62 which is hollow, sealed off at both ends and divided by a central bulkhead 63 into a fore buoyancy chamber 64 and an aft buoyancy chamber 65. The fore buoyancy chamber 64 is equipped with an air tap 66 and a water tap 67. The aft buoyancy chamber is equipped with an air tap 68 and a water tap 69. The taps and associated air hoses are also similar to those provided for the spider legs. Fixed to the boom tube 62 by brackets 71 and 72 is a lead screw 73. The boom tube 62 and its lead screw 73 translate back and forth horizontally in engagement with various rollers 75. This displacement is effected by an electric motor 76 on the frame 53 which, as shown in detail in FIG. 10, drives gears 77, 78 and 79 to move the boom tube 62 telescopically in relation to the boom frame 53. Projecting forwardly from the end of the boom tube 62 is an electric rotator 81 which turns an ultrasonic scanner device 82. A target 83 for calibration of the scanner device 82 is fixed on the frame 53 of the boom assembly. Also affixed on the frame 53 of the boom assembly is a headlamp 84 for use in the event a television monitor is also provided. Suitable pneumatic and hydraulic lines and electrical conductors are provided to the various air and water taps, motors and lights described above. All of these service lines are routed upwardly to the previously mentioned control station on the refuelling bridge 21. The stroke of the scanner 82 on the telescopic boom assembly 54 is almost four feet so that the scanner can move from the retracted position shown in solid lines in FIG. 1 to the dotted line position well within the outlet nozzle 14. The method of the invention with reference to the foregoing apparatus will now be described in relation to FIGS. 11A through 11D. It is to be emphasized that FIGS. 11A-D are wholly schematic and are included solely for purposes of illustrating the sequence of steps carried out during the operation of the apparatus described above. Referring first to FIG. 11 1A, the two spaced vertical guide studs 16 and 17 extend upwardly from the upper circular vessel flange 13, approximately 120 degrees apart with the studs and vessel 10 submerged in a water pool indicated by the surface 20. The scanner 82 is mounted on the horizontal positively driven telescopic boom assembly 54 to which the vertical column 52 is attached. The column is slideable between stop limits 57 and 58 and rotatable within and can be braked with respect to a hub assembly 27 of a spider having three substantially equally spaced legs (24 and 25 of which are shown) adapted to span the flange 13. Clamps 41 and 49 are on the outer ends of the legs 24 and 25. Each of the column 52 and spider legs defines one buoyancy chamber and the boom assembly 54 defines fore and aft buoyancy chambers 64 and 65 into which air and water may be introduced. The method provides for introducing air into each of the various buoyancy chambers in the initial position of the apparatus shown in FIG. 11A. Next the scanner-boom-column-spider assembly is floated in the water pool with the upper end of the column 52 extending above and the scanner 82 and boom assembly 54 extending below the water surface 20. Some water, separate from the pool water, is then introduced into the aft chamber 65 of the boom assembly 54. The brake shoe 32 operating on the column 52 is then turned off and on in short bursts to allow the column 52 to slide gently down through the spider hub assembly 27 until the stop 58 at the upper end of the column 52 is against the spider hub 27 in the position shown in FIG. 11B. A ten-foot section of positioning rod 60 is then attached to the upper end of the column 52 by snapping it into the socket 59. Water is then introduced into all buoyancy chambers until the entire assembly achieves neutral buoyancy. An operator on the refuelling bridge 21 then pushes downwardly on the positioning rod 60, and adds another ten-foot section of positioning rod if necessary. As the assembly is pushed downward it will become more buoyant due to the buoyancy of the upper column section and this is to be counteracted by the introduction of additional water into the buoyancy chambers. The operator then manipulates the rod means and introduces more water (separate from the pool water) into all of the buoyancy chambers so that the two leg clamps 41 and 49 are lowered about the stud guides 17 and 16 respectively and all three spider legs 24, 25 and 26 descend onto the vessel flange 13. The two clamps 41 and 49 are then closed pneumatically to fix the spider relative to the vessel. Next the positioning rod 60 is moved vertically and rotated to bring the scanner to a level coaxial with the nozzle. The boom assembly is then telescopically extended to enable inspection by the scanner within the nozzle and thereafter is retracted. After the inspection, air is introduced into the spider leg chambers to render them positively buoyant while the other chambers are still negatively buoyant. The boom assembly 54 is next rotated by the positioning rod 60 so that it is under the spider leg 26 which does not have a clamp. Clamps 41 and 49 are then opened. Air is introduced first into the boom fore chamber 64 and then into the boom aft chamber 65 so that the entire assembly is buoyant. The rod 60 is then manipulated and eventually detached as the assembly floats upwardly off the studs 16 and 17 to the surface. The braking of the hub 58 on the column 52 is ceased to cause the scanner-boom-column to rise relative to the spider until the column is stopped at its upper limit in the hub. This final position is once again as shown in FIG. 11A. |
abstract | The present invention refers to a method and related plants for removing, by means of redox reactions, the 137Cs from polluted EAF dusts, with an initial average value of radioactivity concentration either higher or lower than 10,000 Bq/kg, the decontamination from the 137Cs initially present in the EAF dusts having a yield of 98%-100%; the present invention also refers to the use of chemical-physical destabilization agents, by means of redox reactions, for obtaining EAF dusts decontaminated from 137Cs. |
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047924292 | claims | 1. In a spring retention cap for a nuclear fuel assembly with an upper end fitting having a leaf spring assembly of a given width and including at least one spring retention fastener, the improvement comprising: a body for engaging the end fitting, two plane exterior surfaces defining a corner of said body, a base transverse to said corner of said body, a slot in said base spaced from and extending substantially parallel to said two plane exterior surfaces, an inwardly directed flange formed by a portion of said base and extending inwardly from said plane exterior surfaces to define a slot opening slightly larger than the given width of said leaf spring assembly, and means for securing said retention cap in position with an end of said leaf spring assembly and a spring retention fastener covered thereby. 2. The spring retention cap of claim 1 in which the end fitting has two plane exterior surfaces forming a corner below the corner of the cap and the flange formed by a portion of the base is dimensioned to be inserted in a slot in each of the end fitting exterior surfaces. 3. The spring retention cap of claim 1 in which a portion of the base on the opposite side of the slot in the base from the flange is inward of the flange such that the flange defines a hook-like structure. 4. The spring retention cap of claim 1 in which the distance from the two spaced exterior surfaces of the body to the slot is sufficient and the means for securing the cap is sufficiently dimensioned to provide strength to accommodate a jacking force created by the leaf spring assembly in the event of failure of the leaf spring assembly retention fastener. 5. The spring retention cap of claim 4 in which two spring retention fasteners are covered by the cap and the jacking force is accommodated in the event of failure of both fasteners. |
abstract | A micro structure which is preferred as an original plate of an antireflection, a mold of nano imprint or injection molding is obtained by a single particle film etching mask on which each particle is precisely aligned and closest packed in two dimensions. A single particle film etching mask is produced by a drip step wherein a dispersed liquid in which particles dispersed in a solvent are dripped onto a liquid surface of a water tank, a single particle film formation step in which a single particle film which consists of the particles by volatizing a solvent is formed, and a transfer step in which the single particle film is transferred to a substrate. The single particle film etching mask on which particles are closest packed in two dimensions, has a misalignment D(%) of an array of the particles that is defined by D(%)=|B−A|×100/A being less than or equal to 10%. However, A is the average diameter of the particles, and B is the average pitch between the particles in the single particle film. |
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043269222 | summary | BACKGROUND OF THE INVENTION 1. Field of the Invention This invention relates to rodded open lattice nuclear fuel assemblies and more particularly provides a fuel assembly having fuel rods of differing diameter. 2. Description of the Prior Art Nuclear reactors typically include a core made up of a plurality of nuclear fuel assemblies disposed within a vessel through which a gaseous or liquid coolant flows. The coolant removes energy in the form of heat as it flows about and through the fuel assemblies. Particularly in pressurized reactors circulating a liquid coolant such as water it is undesirable to reach a boiling condition and, accordingly, reactor cores are designed to prevent a condition commonly referred to as "departure from nucleate boiling" (DNB). If a vapor is permitted to exist along the fuel rods the heat transfer from the heat generating rods to the surrounding coolant is substantially impaired and a potenstial exists for damage to the fuel rods through overheating. Accordingly, DNB related safety criteria impose an upper boundary on the maximum coolant temperature and therefore limit total reactor efficiency. These concerns are further complicated by the desired utilization of mixed oxide or plutonium fuels which are more expensive to fabricate and inherently have a high neutronic capture cross section and a strong moderator temperature coefficient as compared to more commonly used uranium fuels. And, since the coolant becomes progressively hotter as it flows upwardly through the core it therefore progressively changes density, providing progressively less moderation at the upper portion of the core as compared to the bottom. This results in less control and design flexibility. Approaches toward alleviating these limitations, primarily in response to DNB effects, include increasing the water-to-fuel ratio of the core and utilizing full core length fuel rods of smaller cross section which generate less energy per unit length. These approaches, however, are complicated by other factors including excessive fabrication costs and larger cores. Additionally, reactor designs incorporating redundant safety systems which respond to the unlikely event of a rupture of the reactor coolant system causing a loss of coolant, flood the core area with coolant that builds up from the bottom toward the top of the core. Additional flexibility in providing a core design responsive to this condition is advantageous. It is therefore desirable to provide a nuclear fuel assembly which alleviates the temperature and efficiency limitations, and it further is desirable to provide such an assembly which is compatible with mixed-oxide fuel utilization. It is additionally desirable to provide design flexibility to respond to the varying coolant density throughout the core and unlikely coolant discharge occurrences. SUMMARY OF THE INVENTION This invention provides a nuclear fuel assembly which allows higher coolant operating temperatures and enhanced utilization of mixed-oxide fuels. It also provides additional core design flexibility. In one embodiment it provides a vertically oriented assembly of elongated fuel rods of two differing diameters. The upper portion of the assembly includes rods of a smaller diameter than rods of the lower portion of the assembly. Both the smaller upper and the larger lower rods are preferably cylindrical and supported by lattice grid structures which allow axial rod expansion. The grids form cells about each rod for lateral support, and the lowermost grid of the upper rods is preferably directly fixed to the uppermost grid of the lower rods. The grids are also joined through tubular guide thimbles which extend through both the upper and lower rod arrays and through which neutron absorbing control rods can be inserted and withdrawn from the assembly to adjust the neutron flux. A slight space is provided between the upper and lower fuel rods to allow, among other advantageous functions, axial rod expansion without interference among the upper and lower fuel rods. The assembly is advantageously adaptable to use of mixed-oxide fuel in the bottom rods and uranium fuel in the upper rods. In this configuration the space between the upper and lower rods tends to alleviate the sharp power peak which could otherwise occur at the interface of the plutonium and uranium fuels. In a typical reactor the coolant flows upwardly through the core and locating the smaller diameter rods in the upper portion of the assembly produces a lower generated energy per unit length of the fuel rod in the area where the coolant is the hottest. This allows a higher coolant dicharge temperature from the core and a more efficient system. The relative cross sectional size of the individual upper and lower fuel rods can also be varied as dictated by the designer to respond to the change in moderation with coolant temperature change. Further, a plenum is typically provided in the upper portion of most fuel rods to allow accumulation of gaseous fission products released during reactor operation. Because of the lower generation rate in the upper fuel rods, this upper plenum can be of reduced size in comparison to previous fuel rods. A plenum is also provided for the more energetic lower rods; however, it is located at the bottom of the lower rods to reduce flux peaking at the center of the core. Additionally, neutron absorbing materials can be positioned at the top of the lower fuel rods and at the bottom of the upper fuel rods to limit moderation in the area between the two rod arrays. |
description | The present invention relates to the field of nuclear fuel reprocessing and particularly, to a spent fuel dry-process reprocessing method for directly obtaining zirconium alloy fuel. Fuels for fast reactor mainly include two types: an oxide ceramic fuel and a metal fuel. A breeding ratio of the metal fuel used by fast reactor for the purpose of fuel breeding is over 20% higher than that of the oxide ceramic fuel, wherein a zirconium alloy fuel becomes a preferred alloy fuel type for fast reactor due to its excellent nuclear properties and physical properties. Molten salt electrolysis is the most active spent fuel dry reprocessing technology for research. Spent fuel dry reprocessing adopts different molten salt electrolysis processes depending on whether the spent fuel is a metal oxide or a metal spent fuel. At present, the spent fuel dry reprocessing, internationally, mainly adopts a chloride molten salt electrolysis refining technology which is as follow: for the metal oxide spent fuel, metal oxide is first reduced to metal using metal lithium (or metal lithium is generated by electrolysis), and then the obtained metal, on a chloride molten salt, is subjected to controlled potential melting and controlled potential deposition, so as to respectively obtain uranium and uranium plutonium alloy fuel. A patent reported a dry reprocessing molten salt system for preparing an aluminum alloy fuel. Aluminum alloy cannot completely meet the needs of the fast neutron reactor due to its low melting point. However, the zirconium alloy fuel has a high melting point, which is beneficial to the stable operation of the fast reactor. At present, the method for preparing the zirconium alloy fuel internationally comprises the steps of first obtaining uranium, plutonium and zirconium metals and then mixing the three metals in a ratio for melting. But, a process of preparing an alloy fuel by melting with pure metals is extremely high in cost, in which pure metal materials are first obtained through a metallurgical process and then the pure metals in a ratio are mixed for melting, and thus, the process is complicated, waste is high, energy consumption is high and cost is large. With respect to the defects existing in the prior art, it is an object of the present invention to provide a spent fuel dry reprocessing method for directly obtaining the zirconium alloy fuel, and this method is simple and controllable in process, low in energy consumption, low in cost and easy to industrialize. In order to achieve the above object, a technical solution adopted by the present invention is as follows: a spent fuel dry reprocessing method for directly obtaining zirconium alloy fuel comprises the following steps: determining components and a ratio of a molten salt composition used for melting a spent fuel according to a zirconium alloy fuel design requirement and contents of actinium series metals in the spent fuel; melting the spent fuel in the above molten salt composition; and selecting an electrode pair for electrodeposition so that zirconium in the molten salt composition and uranium ions in the spent fuel or uranium and other actinium series metal ions are subjected to co-deposition, thereby obtaining a zirconium alloy nuclear fuel meeting the zirconium alloy fuel design requirement. Further, the molten salt composition at least comprises zirconium fluoride, potassium fluoride and lithium fluoride, and a mole ratio of zirconium fluoride to potassium fluoride to lithium fluoride is 1:(10-20):(25-80). Further, determining components and a ratio for a molten salt composition comprises the steps: first, determining an amount of zirconium fluoride according to a zirconium alloy nuclear fuel design requirement, and then determining amounts of potassium fluoride and lithium fluoride according to a Nernst equation and a target melting point temperature of the molten salt composition. Further, a mixture of the spent fuel and the molten salt composition is molten at 600-1050° C., and a preferred melting temperature is the lowest eutectic point temperature +50° C. of the proportioned molten salt. Further, according to the zirconium alloy fuel design requirement, if some metal ions in the molten spent fuel are in excess, pre-electrodeposition is carried out through an additional electrode pair so that the excess metal ions in the spent fuel are separated out. Further, if the spent fuel is a metal oxide spent fuel, a positive electrode of the electrode pair used in the pre-electrodeposition processes to separate out the excess metal ions and a positive electrode of the electrode pair used in electrodeposition to form a zirconium alloy nuclear fuel is made of inert material. Alternatively, if the spent fuel is a metal spent fuel, the positive electrode of the electrode pair used in the process of pre-electrodepositing to separate out the excess metal ions is the metal spent fuel. The method provided by the present invention is simple and controllable in process, there is no a process of preparing pure metals but the spent fuel is directly added into the molten salt composition, and the desired zirconium alloy fuel is directly obtained through electrodeposition as long as the ratio of the molten salt composition is adjusted; in addition, this method is suitable for a metal oxide spent fuel and a metal spent fuel, and is low in energy consumption, low in cost and easy to industrialize. The present invention will be further described in combination with embodiments. wt % in examples represents a mass percent. This example adopts an uranium dioxide ceramic spent fuel. A target alloy fuel: a zirconium-uranium alloy fuel, which comprises the components: 95 wt % of uranium and 5 wt % of zirconium. A preparation method of the zirconium-uranium alloy fuel is as follows: determining a molten salt composition: 5 wt % of zirconium fluoride, 35 wt % of potassium fluoride and 60 wt % of lithium fluoride according to the components of the above zirconium-uranium alloy fuel, components of the spent fuel and an electrodeposition principle (Nernst equation); directly putting 11 g of sintered uranium dioxide pellet into 1500 g of molten salt containing 5 wt % of zirconium fluoride, 35 wt % of potassium fluoride and 60 wt % of lithium fluoride, and melting the uranium dioxide pellet in the molten salt composition at a temperature of 750° C.; and carrying out pre-electrodeposition using an electrode pair with graphite as a positive pole and an uranium bar as a negative pole, wherein the controlled potential of the negative pole is −1.2˜−1.4V (relative to a Ag/AgCl reference electrode) so that 1.2 g of excess uranium is deposited on the negative pole; and subsequently, carrying out electrodeposition using an electrode pair with graphite as the positive pole and a metal uranium wire having a diameter of 1 mm as the negative pole, wherein the controlled precipitation potential of the negative pole is −1.6˜−2.0V (which is adjusted according to a uranium-zirconium ratio of deposited alloy and a requirement on a decontamination factor of rare earth) so that zirconium is first deposited on the negative pole, then uranium is also deposited on the negative pole through an under-potential deposition principle to form alloy with zirconium, and finally, 95 wt % uranium-5 wt % zirconium alloy in a calculated ratio is obtained on the negative pole. This example adopts an uranium plutonium alloy spent fuel. A target alloy fuel: a zirconium-uranium-plutonium alloy fuel, which comprises the components: 76 wt % of uranium, 5 wt % of zirconium and 19 wt % of plutonium. A preparation method of the zirconium-uranium-plutonium alloy fuel is as follows: determining a molten salt composition: 10 wt % of zirconium fluoride, 50 wt % of potassium fluoride and 40 wt % of lithium fluoride according to the components of the above zirconium-uranium-plutonium alloy fuel, components of the spent fuel and an electrodeposition principle (Nernst equation); directly putting 50 g of uranium-plutonium alloy spent fuel as a positive pole into 3000 g of molten salt containing 10 wt % of zirconium fluoride, 50 wt % of potassium fluoride and 40 wt % of lithium fluoride, and carrying out controlled potential melting and controlled potential electrodeposition at a temperature of 850° C.; and when the controlled melting potential is 1.2-1.6V (relative to a Ag/AgCl reference electrode), melting a uranium-plutonium alloy spent fuel in the above molten salt composition; carrying out electrodeposition using double negative poles, wherein both of negative pole materials use metal uranium wires having a diameter of 1 mm, the controlled potential of one of the negative poles is −1.2˜−1.5V (relative to a Ag/AgCl reference electrode) so that 21 g of excess uranium is deposited on the negative pole, the controlled precipitation potential of the other negative pole is −1.8˜−2.3V (which is adjusted according to a zirconium-uranium-plutonium ratio of a target deposited alloy and a requirement on a decontamination factor of rare earth) so that zirconium is first deposited on the negative pole, then uranium is deposited on the negative pole, and then plutonium is also deposited on the negative pole through an under-potential deposition principle to form alloy with zirconium, and finally, 76 wt % uranium-5 wt % zirconium-19 wt % plutonium alloy in a calculated ratio is obtained on the negative poles. By adoption of the method provided by the present invention, the concentration of zirconium fluoride in the molten salt and the concentration of uranium remained in the molten salt after pre-electrodeposition are adjusted so as to obtain uranium-zirconium alloy having a (uranium+plutonium)-zirconium mole ratio of 100-50. The above examples merely exemplify the present invention, and the present invention can also be implemented in other specific manners or other specific forms without departing from the essence or substantive characteristics of the invention. Thus, the described embodiments are all considered as being illustrative but not limiting in term of any viewpoints. The scope of the invention should be described by additional claims, and any variations equivalent to the intention and scope of the claims should be contained in the scope of the present invention as well. |
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abstract | The pattern observation method for observing a pattern which is formed on an insulating film, includes: irradiating an entirety of the pattern with a charged particle beam, to obtain a temporary image of the pattern which has region information of a convex pattern and a concave pattern; irradiating the convex and concave patterns with the charged particle beam having a first and second voltages based on the region information, to thereby form an electric field between a top surface of the convex pattern and a bottom surface of the concave pattern so that charged particles emitted from the bottom surface of the concave pattern may be drawn out to an outside of the pattern; and irradiating the entirety of the pattern with the charged particle beam to obtain an image of the pattern having the information of the bottom surface of the concave pattern. |
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description | The invention relates to an Anti-Scatter-Grid for a radiation detector, to a radiation detector with such an Anti-Scatter-Grid, to an examination apparatus with such a detector, and to a method for the determination of scattered radiation impinging on a radiation detector. During the generation of images by the spatially resolved measurement of radiation intensity it is often necessary or desirable to discriminate radiation coming from a target direction from radiation coming from other directions. A typical example of this situation is the generation of medical X-ray images by a CT or some similar device. In this case the main interest is the spatially resolved measurement of primary radiation coming directly from the X-ray source. Due to scatter processes in the examined body, however, there is always secondary or scattered radiation present that impinges on the detector from directions other than the line of sight to the X-ray source, wherein such scattered radiation carries no usable image information. In order to improve image quality and the signal-to-noise ratio, it is therefore known to dispose an Anti-Scatter-Grid (ASG) in front of the detector, wherein said grid comprises radiation absorbing lamellae which define channels through which radiation from a target direction may freely pass, while scattered radiation coming from other directions will be largely absorbed. It is however impossible to remove all scattered radiation with an Anti-Scatter-Grid, and therefore the contribution of scattered radiation to a measured image signal remains a problem, particularly in image regions of weak signal intensity. Though different approaches have been tested to deal with the aforementioned problems, no satisfying solution could yet be found. The U.S. Pat. No. 6,618,466 B1 proposes for example a method in which the contribution of scattered radiation to an image is determined via the generation of images with and without a beam stop array and the application of interpolation techniques. Such a laborious method is however hardly to integrate into the normal medical workflow. Based on this situation it was an object of the present invention to provide means for the reduction of negative effects of scattered radiation on image quality. This object is achieved by an Anti-Scatter-Grid, a radiation detector, an examination apparatus and a method. According to its first aspect, the invention relates to an Anti-Scatter-Grid for a radiation detector which comprises lamellae that strongly absorb radiation of a certain spectrum which shall be observed by the detector, for example X-radiation. The radiation may pass freely only through (void or transparent) channels which are established between the lamellae. The channels thus define lines of sight along which radiation may reach a detector at the “backside” of the Anti-Scatter-Grid. The Anti-Scatter-Grid may be one-dimensional, wherein the channels are planes between parallel lamellae, or two-dimensional, wherein two sets of parallel lamellae cross and define channels. Depending on the intended application of the Anti-Scatter-Grid, the channels/lamellae may for example be parallel to each other or focused on a certain point (typically a radiation source). Moreover, it is an important feature of the radiation absorbing lamellae that they are adapted to produce a signal that indicates the amount of radiation absorbed by said lamellae. The signal may for example be proportional to the total energy of the absorbed radiation. An Anti-Scatter-Grid of the aforementioned kind absorbs scattered radiation coming from directions other than a target direction defined by the channels of the grid, thus shielding a detector behind the Anti-Scatter-Grid from undesired radiation. Besides this, the Anti-Scatter-Grid yields a signal that indicates the amount of absorbed radiation. Based on this signal it is possible to estimate the fraction of scattered radiation that is actually present and thus also the amount of scattered radiation that reaches the detector despite of the Anti-Scatter-Grid. This in turn allows to correct the image signals of the detector and to improve image quality. According to a preferred embodiment, the lamellae of the Anti-Scatter-Grid comprise a semiconductor material that converts absorbed radiation into electrical signals. The electrical signals may particularly be based on the generation of free current carriers (e.g. electron-hole pairs). In order to maximize the sensitivity of the aforementioned semiconductor lamellae, it is preferred that they comprise a material with a low intrinsic energy conversion coefficient for the conversion of photons of the absorbed radiation into electron-hole pairs, wherein said coefficient may particularly be lower than 10 eV per electron-hole pair. According to another embodiment of the invention, the lamellae comprise a scintillator material for the conversion of incident radiation of a first energy level (e.g. X-rays) into radiation of a second energy level (e.g. visible photons). The second energy level of the radiation generated in the scintillator material is then typically such that it may more readily be detected, e.g. by the aforementioned semiconductor material. The scintillator material is preferably disposed as an outer layer on the surface of the lamellae. Moreover, the lamellae may preferably comprise a material with a high absorption coefficient, particularly higher than 1 cm−1, for photons with energies below 150 keV. Such lamellae make the Anti-Scatter-Grid suited for use with an X-ray detector. The material with such a high absorption coefficient may particularly be a heavy metal with an atomic weight above 40. One preferred example of a material for the lamellae is CdZnTe which has both a low intrinsic energy conversion coefficient and a high absorption coefficient for X-rays. As was already mentioned, the lamellae may particularly generate an electrical signal that corresponds to the dose of absorbed radiation. In this case, the lamellae are preferably covered completely or partially by electrodes, said electrodes allowing to induce an electrical field inside the lamellae and to collect charge carriers generated by absorbed radiation. The electrodes may typically consist of metal, particularly of Pt. Preferably there is a plurality of separate electrodes which allow a spatially resolved measurement of absorbed radiation. In a two-dimensional Anti-Scatter-Grid, one electrode may for example be located on each wall of each channel. In the aforementioned embodiment, at least one of the electrodes may optionally end a distance away from the edge of the corresponding lamella on which said electrode resides. Processes in the margin of said lamella are therefore not influenced by the electrode. The distance of the electrode from the edge is preferably so large that radiation will substantially (e.g. to more than 90%, preferably more than 98%) be absorbed by the material of the lamella over said distance. Therefore, (primary) radiation that is parallel to the lamellae and hits them at their edge will be absorbed in the lamella within the margin not covered by the electrode; contributions from said radiation will thus not be measured by the electrode, allowing the better separation between primary and scattered radiation. More details on this topic will be discussed in the description of preferred embodiments. The aforementioned electrodes which end a distance away from the edge are preferably mixed in the Anti-Scatter-Grid with electrodes that extend to the edge. The long and short electrodes may for example alternate from pixel to pixel. Such an arrangement allows a spatially alternating measurement of radiation absorbed by the lamellae with and without contributions from primary radiation. Therefore, the fractions of primary radiation and scattered radiation can be discriminated in a spatially resolved way. The invention further comprises a radiation detector with the following components: An Anti-Scatter-Grid of the aforementioned kind which comprises radiation absorbing lamellae that produce a signal related to the amount of absorbed radiation. Optionally a converter for the conversion of incident radiation of a first energy level into radiation of different energy level. In X-ray detectors, such a converter is known as scintillator which converts incident X-rays into photons of visible light. An array of radiation sensitive sensor units that can measure the radiation passing through the Anti-Scatter-Grid (directly or after conversion by the converter) in a spatially resolved way. A signal processing unit for the evaluation of signals generated by the Anti-Scatter-Grid. The signal processing unit may particularly comprise electronics for the amplification, correction and/or analogue-to-digital conversion of signals from the Anti-Scatter-Grid. Moreover, it may comprise a digital data processing device (with e.g. microprocessor, memory, I/O interfaces etc.) for the evaluation of said signals. According to a preferred embodiment, the signal processing unit may be adapted to discriminate fractions of incident radiation with respect to their parallelism to the channels/lamellae of the Anti-Scatter-Grid. Thus, the signal processing unit may (at least approximately) determine the fraction of the incident radiation that is parallel to the channels and therefore probably primary radiation as well as the fraction of radiation that is not parallel to the channels and therefore probably scattered radiation. Preferably, the processing unit determines said fractions spatially resolved in order to allow a local correction of an image generated by the sensor units. The invention further relates to an examination apparatus, particularly an X-ray, CT (Computed Tomography), PET (Positron Emission Tomography), SPECT (Single Photon Emission Computed Tomography) or nuclear imaging device, which comprises an X-ray sensitive radiation detector of the aforementioned kind. Finally, the invention comprises a method for the determination of scattered radiation impinging on a radiation detector with an Anti-Scatter-Grid, wherein the amount of radiation absorbed by the Anti-Scatter-Grid is directly measured. The measurement is preferably executed in a spatially resolved way. The radiation detector, the examination apparatus and the method share the essential features of an Anti-Scatter-Grid of the kind described above. Therefore, reference is made to the preceding description for more information on the details, advantages and improvements of these objects. These and other aspects of the invention will be apparent from and elucidated with reference to the embodiment(s) described hereinafter. FIG. 1 depicts in a principal sketch (not to scale) a section through a (small part of a whole) detector array 5 of single sensor units 9 (pixels) that is disposed underneath a two-dimensional Anti-Scatter-Grid ASG. Such an arrangement may for example be used in a CT device for the spatially resolved measurement of X-radiation. The X-ray signal of the sensor units 9 is then a superposition of both the transmitted primary radiation 7 (the signal one is interested in) and a signal arising from scatter radiation 8 (which is undesirable and reduced by the Anti Scatter Grid ASG). The presence of scattered X-rays 8 in the projections of a CT acquisition leads to cupping and shadowing artifacts, and thus visible degradation of the reconstructed CT image. In the region near the detector centre, the scatter radiation 8 is typically of a similar order of magnitude as the primary radiation 7, and might even cover the primary signal in areas of weak intensity. Typically, 5% to 15% of the incident scatter radiation 8 are transmitted through an ASG used in front of the detector 5. However, especially for areas with high attenuation, the knowledge of the primary intensity is crucial for image reconstruction. Presently, scatter radiation 8 is detected—if at all—by extra detector cells outside the active detector area of a CT system. The scatter signal within the detector area is determined by interpolation from the values measured at the border. The quality of the interpolation is—more or less—sufficient for small fan beam width, however, it does not take into account shadowing of scatter radiation by small objects within the fan beam. For large area detectors, this concept fails, since the distance between the border detectors will not give detailed information about the scatter distribution in the centre detector area. The main problem outlined in the previous section is the unknown detailed distribution of scattered photons 8 in the measured tomographic projections leading to image degradation. This effect becomes even more important in future systems, since the trend towards large scan areas (cone beam) increases the amount of scatter radiation rapidly due to a larger irradiated area. In the following an approach is described that gives access to a direct measurement of the scatter distribution for the whole detector area with a spatial resolution within the range of the pixel size. The basic idea of this approach is the detection/measurement of scattered X-rays 8 by the ASG itself. The measurement gives information on the spatial scatter photon distribution, allows for an improved scatter correction in the measured projections, and thus results in a higher CT image quality. According to the preferred embodiment shown in the Figures, the Anti-Scatter-Grid ASG comprises walls or lamellae 2 of a direct conversion material that detects scattered photons 8 (“active detection”) and simultaneously absorbs these quanta analogous to a conventional ASG (“passive collimation”)—which means that the ASG acts as an “active collimator”. The direct converting material may for example be a (crystalline, polycrystalline or amorphous) semiconductor which converts the energy of absorbed radiation 1, 8 directly into electron-hole pairs. Preferably the material has both a low intrinsic energy conversion coefficient (e.g. CdZnTe: 5 eV/electron-hole pair) resulting in a high response signal, and a high absorption coefficient for the relevant energy range (>30 keV). In addition, the direct conversion material should have a high atomic weight Z to absorb X-ray photons most effectively (e.g. for CdZnTe: Z≈49). In its geometrical dimensions the grid ASG is similar to a conventional two-dimensional Anti-Scatter-Grid with a typical thickness of the lamellae 2 of approximately 100 μm. The height is chosen appropriately concerning absorption efficiency, technical availability, and costs with a typical range being 10 mm to 50 mm. Between the lamellae 2, a channel is formed through which radiation 7, 8 may pass to sensor units 9 or pixels located underneath the ASG in a detector array 5. The lamellae 2 are covered on each side by an electrode 3, 4, 6, wherein the opposing electrodes are driven as anode and cathode, respectively, by an external circuit (only schematically shown in FIG. 2 for two electrodes). The electrodes 3, 4, 6 at the sidewalls consist of a very thin layer of metal (e.g. Pt) each. The metal layer must be thin enough to ensure that only a negligible amount of scattered X-rays 8 is absorbed without giving rise to a signal. Normally, an electron photocurrent through the direct conversion material of the lamellae 2 is the preferred mode of operation, since the mobility-lifetime product for electrons is much higher than for holes (e.g. CdZnTe (optimized material): μeτe≈3·10−3 cm2/V>>μhτh5·10−5 cm2/V). Typically, the applied electric fields are of the order of 1 kV/cm. So, for a 100 μm thick lamella a bias voltage of only 10 V is needed to ensure sufficient charge collection efficiency. This low bias voltage involved in the operation of the detector lamellae 2 is easy to handle and makes no special protection of circuits necessary. With the help of the electromagnetic field (bias field) that is induced via the electrodes 3, 4 and 6, charge carriers are separated and collected by drift to the electrodes giving rise to an electric signal which can be processed by succeeding readout electronics 10 (FIG. 2). For each rectangular pixel 9 this signal can be considered as an integral value of the four lamellae surrounding it. Thus, a spatial intensity distribution of the scatter radiation 8 in the detector array 5 is obtained with the resolution of one pixel. Concerning the geometry of the electrodes there are two concepts shown in FIG. 1. In a first concept the electrode 4 (preferably the anode contact) covers the complete height of the ASG up to the upper edge of the lamella 2. This results in a sensitivity of the direct conversion material 2 not only to scatter radiation 8 but also to primary radiation 1 which impinges on the upper end face of the lamella 2, causing some difficulties to disentangle the real scatter distribution afterwards. In contrast, electrodes 6 (anode contacts) ending a distance d short below the top edge of the lamella 2 are insensitive to radiation impinging on top of the ASG. Thus, these electrodes 6 measure the whole scatter radiation 8 distribution with the exception of the distribution from the direction of the primary X-ray beam 1, i.e. they are highly sensitive to scatter radiation 8. To ensure a total absorption of at least 99% of X-ray radiation with energies below 120 keV one must allow for an absorption layer of a certain thickness d (e.g. of d=6 mm for CdZnTe). This means that the short anode 6 should be at least 6 mm shorter than the long electrodes 3, 4. By this, one lowers the electric fields in the top region of the lamella 2 to such an extent that charge generated by the absorption process in this region has no contribution to the photocurrent, since it cannot be collected by the short anode 6. As indicated in the top view of the Anti-Scatter-Grid ASG in FIG. 3, one can combine the concepts of long electrodes 3, 4 and short electrodes 6 leading to a pattern like a chessboard which allows to acquire different primary-to-scatter ratios (i.e. lamellae covered with long electrodes 3, 4 only measure full primary and scatter intensity, and lamellae covered with one short electrode 6 measure mainly scatter radiation). As will be discussed below, it is thus possible to precisely determine and separate the primary signal from the mixed primary/scatter signal measured at the pixilated detector. Alternatively, one can use each concept alone (cf. FIG. 4 for long electrodes 3, 4) which allows also to correct for the signal of each detector pixel 9, but with a higher systematic error due to the fact that the radiation distribution from different sources cannot be disentangled. Any detector pixel n of the detector gives a signal In which is a linear combination of incident primary dose Pn (radiation 1+7 in FIG. 1) and incident scatter dose Sn (radiation 8 in FIG. 1). In the following it is assumed that only scatter radiation nearly perpendicular to the surface of the detector 5 (≈90°) can pass the ASG lamellae 2 without being absorbed, i.e. Sn=Sn (≈90°):In=αn·Pn+βn·Sn(≈90°). (1) Scatter radiation with angles different from a small interval around 90° is strongly absorbed by the ASG. Lamellae 2 with long anodes 4 up to the top edge of the ASG lead to a signal I′n per pixel which is given by the following equation:I′n=α′n·Pn+β′n·Sn(≈90°)+γ′n·Sn(≈90°) (2) The intensity signal I″n per pixel obtained from lamellae 2 with short anodes 6 is nearly insensitive to both primary radiation and scatter radiation with ≈90°:I″n=γ″n·Sn(≠90°). (3) The signals sampled from the scatter detector arrays have to be calibrated and corrected. The coefficients an αn, βn, γn, α′n, β′n, γ′n, and γ″n can be determined by an appropriate calibration procedure. In standard CT detectors, the scintillator pixels 9 below the ASG, cognizant for detection of primary radiation 7, will ideally have coefficients αn→1 and βn→0 (cf. Eq. (1)). The scatter detector values I′n including radiation impinging on top of the ASG (cf. Eq. (2)) will ideally have coefficients α′n→0 and β′n→1 and γ′n→0. In this ideal case the signal I″n (cf. Eq. (3)) would be irrelevant to determine Pn per pixel. In the general case of a real detector setup, however, the relevant scatter corrected signal Pn per pixel is obtained algebraically by solving the system of linear equations (1)-(3): P n = β n ′ γ n ″ I n - β n γ n ″ I n ′ + β n γ n ′ I n ″ α n β n ′ γ n ″ - α n ′ β n γ n ″ ( 4 ) The signal intensities In, I′n and I″n are measured and the other parameters are known by a dedicated calibration procedure. In summary, the following advantages of the proposed ASG detector modules based on the principle of active collimation can be achieved: Improved CT image quality by scatter distribution measurement and subsequent scatter correction in CT systems. The scatter distribution is measured for the complete detector area, not only for certain punctual areas. The Anti Scatter Grid acts simultaneously as X-ray detector and as absorber (active collimation) leading to a very compact and simple setup. Since direct conversion material might be also the choice for CT detectors it is advantageous with respect to costs and technical feasibilities to have an ASG made from the same material as the detector pixels. No sophisticated optical projection system is necessary. Thus, there is neither a distortion in geometry as well as in intensity distribution of the detected scatter, nor difficulties concerning the signal-to-noise ratio. Since the ASG itself is used as scatter detector, there is only additional but never any loss in information and performance of the CT detector and thus in image quality. Selective detection of scatter radiation—which is done by a special combination of longer and shorter electrodes making some pixels insensitive for primary radiation impinging at a lamella on top—gives a very precise access to the relevant primary radiation. Finally it is pointed out that in the present application the term “comprising” does not exclude other elements or steps, that “a” or “an” does not exclude a plurality, and that a single processor or other unit may fulfill the functions of several means. Moreover, reference signs in the claims shall not be construed as limiting their scope. |
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abstract | A mobile heat pipe cooled fast nuclear reactor may be configured for transportation to remote locations and may be able to provide 0.5 to 2 megawatts of power. The mobile heat pipe cooled fast reactor may contain a plurality of heat pipes that are proximate to a plurality of fuel pins inside the reactor. The plurality of heat pipes may extend out of the reactor. The reactor may be configured to be placed in a standard shipping container, and may further be configured to be contained within a cask and attached to a skid for easier transportation. |
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048250902 | abstract | An EMI/RFI radiation and static electricity suppression device for shielding radiation from various electronic and electrical sources. The suppression device includes a shielding membrane having the waveguide barrier with a multiplicity of protrusions facing the radiation emitting source for minimizing reflection of electromagnetic radiation incident thereupon. A plurality of conductive layers are each separated by spacers formed of a radiation absorbing media. The conductive layers include an inner-mesh disposed adjacent to the waveguide barrier and adapted for blocking low frequency radiation. A deflection layer formed of a multiplicity of spaced apart metallic deflection pads is provided for deflecting incident radiation thereabout. An absorption funneling layer disposed behind the deflection layer includes a plurality of apertures designed to receive the deflected radiation. A reflecting plate disposed behind the absorption funneling layer works in combination with the absorption funneling layer to trap radiation and cause its ultimate absorption in an absorptive region therein. Each of the conductive layers is connected to a common electrical ground. Impinging electromagnetic radiation emanating from a device may be effectively shielded with such a construction and electrostatic field build-ups are eliminated. |
060463740 | claims | 1. A cellular cement grout for forming a radiation barrier around a radioactive component, said grout comprising in mixture: a cement slurry; a finished foam material; a gel-forming material; and a radiation-absorbing metal constituent. a cement slurry; a finished foam material; a bentonite gel material; and a radiation-absorbing metal constituent. a container which encloses said radioactive component; and a cellular cement grout which forms a radiation barrier in said container around said radioactive component, said grout comprising in mixture: 2. The grout of claim 1, wherein said gel-forming material comprises a bentonite gel material. 3. The grout of claim 1, wherein said radiation-absorbing metal constituent comprises an iron-containing material. 4. The grout of claim 1, wherein said radiation-absorbing metal constituent comprises a barium-containing material. 5. The grout of claim 1, wherein said radiation-absorbing metal constituent comprises a rounded metal shot material. 6. The grout of claim 5, wherein said rounded metal shot material comprises rounded steel shot material. 7. The grout of claim 6, wherein said rounded steel shot material comprises steel shot in the size of approximately size 2 to size 4 bird shot. 8. A cellular cement grout for forming a radiation barrier around a radioactive component, said grout comprising in mixture: 9. A radiation barrier for a radioactive component, said barrier comprising: 10. The barrier of claim 9, wherein said gel-forming material comprises a bentonite gel material. 11. The barrier of claim 9, wherein said radiation-absorbing metal constituent comprises an iron-containing material. 12. The barrier of claim 9, wherein said radiation-absorbing metal constituent comprises a barium-containing material. 13. The barrier of claim 9, wherein said radiation-absorbing metal constituent comprises a rounded metal shot material. 14. The barrier of claim 13, wherein said rounded metal shot material comprises rounded steel shot material. 15. The barrier of claim 14, wherein said rounded steel shot material comprises steel shot in the size of approximately size 2 to size 4 bird shot. |
abstract | An emergency core cooling system is provided with a hybrid safety system composed of an active safety system and a static safety system for ensuring the safety against a severe natural phenomenon such as a giant earthquake and a mega hurricane. An emergency core cooling system for a boiling water reactor includes four safety divisions in total: three safety divisions for an active safety system having a high pressure reactor core cooling system, a low pressure reactor core cooling system, a residual heat removal system, and an emergency diesel generator; and one safety division for a static safety system having an isolation condenser, a gravity drop reactor core cooling system, and a static containment vessel cooling system. |
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054597684 | description | DESCRIPTION OF THE PREFERRED EMBODIMENTS Referring now to the figures of the drawing in detail and first, particularly, to FIG. 1 thereof, there is seen a diagrammatic, sectional view of a nuclear reactor pressure vessel 1 (referred to below as a pressure vessel) of a pressurized-water nuclear power station which is constructed, e.g., for a thermal reactor output of 3765 MW, corresponding to a gross electrical output of 1300 MW. A reactor core 2, which is composed of fuel assemblies of which only a single one 3 is shown, is cooled with light water which enters through inlet ports 4 and flows downward in an annular cavity 5 (shown by flow arrows f1). The cooling water flows upward from a bottom plenum 6 through a perforated lower grid 7, through cooling ducts of the fuel assemblies 3 in which it is warmed, and then flows from an upper plenum 8 through outlet ports 9 and a so-called hot primary-circuit pipe or pipeline 10 connected thereto to a non-illustrated steam generator where it conveys its heat through heat-exchanging tubes to a secondary coolant. The cooling-water flow through the reactor core 2, the upper plenum 8 and the outlet ports 9 is illustrated by flow arrows f2. The cooled cooling water, which is also known as the primary coolant, is pumped back from the steam generator through the non-illustrated so-called cold primary-circuit pipe to the inlet port 4 of the pressure vessel 1, so that a continuous circulation is established in normal operation. In normal operation, the primary coolant which is in the primary circuit and therefore is also inside the pressure vessel 1, is at a pressure of approximately 158 bar, and coolant temperature at the outlet port 9 is approximately 329.degree. C. The reactor pressure vessel 1, with its fittings, is constructed for this pressure and temperature load, including a safety margin. It is formed of a cup-shaped vessel bottom section 1A with a hemispherical dome 11 and a flange ring 12 at its upper end, to which a domed cover 1B having a counter flange 13 is bolted in a sealing manner. Cover bolts which are used are not shown, but bolt passage holes 14 can be seen. Only the most important of the fittings will be mentioned, which are a lower perforated drum 15 and the previously mentioned lower grid 7 above the lower perforated drum 15 which forms a bottom of a core barrel 16. The core barrel 16 is suspended by means of a supporting flange 16.1 on an annular shoulder 17 of the flange ring 12 and has a bottom section in which it accommodates the core 2 with the individual fuel assemblies 3. The core 2 is covered by an upper grid plate 18 on which a guide framework 19 having an upper support plate 19.1 is supported. Control rods 20 which can be lowered or raised by non-illustrated control rod drives disposed above the cover 1B, are inserted into a portion of the fuel assemblies. Four outlet ports 9 and four inlet ports 4 are alternately distributed over the perimeter of the pressure vessel 1 in a plane 21--21, in a four-loop system. The primary coolant, which is held below a supercritical pressure and is therefore liquid in normal operation, not only covers the core 2 but also fills the upper plenum 8 up to about the upper support plate 19.1. Effective cooling is therefore ensured even of those fittings which are subject to so-called gamma heating by virtue of gamma radiation, even though they do not themselves generate heat (such as the fuel assemblies 3). If the water level in the pressure vessel drops, due to an extremely improbable failure of all of the cooling and emergency cooling devices, the temperature of the assembly (normally approximately 400.degree. C.) starts to rise, and heat is increasingly conveyed, especially by radiation and conductance, through the pressure vessel 1, especially if the water level has dropped to the upper grid plate 18 or even slightly below. This overheating is utilized, in the still relatively early stage, by the safety device according to the invention, in order to reliably prevent overpressure failure of the pressure vessel 1 in the case of an inadequate core cooling mentioned above. To this end, a differential pressure-loaded pressure relief valve 22 is set into a coolant conducting surface or conduit which may be a wall of the pressure vessel 1 or a pipeline connected to the pressure vessel (the conduit shown is the hot primary coolant pipe 10). As is seen in FIG. 2, the valve 22 has a closure piece that is mounted in such a way as to be longitudinally displaceable. The closure piece is preferably a differential-pressure piston 23 which is held by a fusible stop 24, 25 in an illustrated closure position thereof so as to provide a seal. When the reactor interior reaches an upper threshold temperature which, for example, is 700.degree. C., the fusible stop 24, 25 is caused to melt due to a threshold temperature heat flow reaching it. It can then no longer withstand the shear forces acting on it, so that the differential-pressure piston 23 is displaced into its open position due to the differential-pressure forces acting on it. The pressure difference is calculated as: P=P.sub.1 -P.sub.0, where P.sub.1 =internal pressure, and P.sub.0 =external pressure. The pressure P.sub.1 prevailing in the interior of the pressure vessel 1 can thus be reduced through a connection pipe socket 26 and an open cross section of the annular duct 27 leading into a pressurizer discharge line 28, through a pressure relief flow shown by arrows f3. The fusible stop 24 is disposed between sealing surfaces 23.1 of the piston 23 and associated seating surfaces 29 of the pressure relief valve 22. In the illustrated example, these seating surfaces 29 are formed by internal peripheral surfaces of the connection pipe socket 26 which may be provided with an additional reinforcement in the region of the seating surfaces. The fusible stop is constructed in such a way that at normal temperature it can easily withstand a differential pressure of 160 bar. As an additional safeguard, the further fusible stop 25 is disposed between peripheral piston surfaces 23.2 (at an insertion end of the piston 23) and guide surfaces 30 at an inner periphery of a guide cylinder 31. The piston 23 and the guide cylinder 31 are preferably constructed as hollow bodies, because such a structure causes the heat flow to reach the fusible solder selectively and without major losses. The guide cylinder 31 is held in a centered position in a valve body 220 in such a way that the annular cavity or annular duct 27 remains free as an overflow duct, between the outer periphery of the guide cylinder 31 and the inner periphery of the valve body 220. Vanes 32 which are disposed in the annular cavity 27 hold the guide cylinder 31 in a centered position and are joined to a wall of the valve body. As is shown, the inlet cross section of the annular cavity or overflow duct 2 is sealed by the piston 23 in the normal position of the piston, but is cleared in a release position thereof. In the release position, the piston 23 is completely inserted into the guide cylinder 31. In order to generate the differential pressure at the piston 23, and in order to facilitate the insertion movement, the guide cylinder 31 has a bottom 33 which is provided with a pressure relief orifice 34 therein. As was already mentioned and as is shown, the pressure relief valve 22 is set into the wall of the primary coolant pipe 10, and specifically into the wall of the so-called hot primary circuit, close to the pressure vessel 1. According to an alternative embodiment, the connection pipe socket 26 of the pressure relief valve 22 could be set into the cylinder wall of the pressure vessel 1 at the level of the primary coolant pipe sockets 4, 9 which are shown in FIG. 1, and specifically in a non-illustrated circumferential interspace therebetween. Alternatively, the pressure relief valve 22 may be constructed as a pressure control valve 22' shown in FIG. 3. The pressure control valve 22' has correspondingly smaller cross-sectional dimensions, and instead of the pressure control line 28, a pressure control line 28' is provided which is connected to a servo piston unit 35 of a relief valve 36. This relief valve 36 is connected on the input side to the system pressure P.sub.1, which it normally shuts off from a line 37 running to a non-illustrated pressurizer relief tank. It is only if the control valve 22' responds, in the case of overheating of the pressure vessel 1, that the relief valve 36 be would opened. This relief valve may be the relief valve which in any case is connected to the primary circuit in the vicinity of the pressurizer in conventional nuclear reactor plants. The valve 22' (in its smaller embodiment as a pressure control valve) may alternatively be installed in the interior of the pressure vessel 1, e.g., in the vicinity of the lower grid 7 or the upper grid plate 18, so that it is disposed even closer to possible hot spots and thus responds even more quickly. The associated pressure control line 28' would then be run to the outside, in a pressure-tight manner, in the form of a thin measuring line through the cover 1B or at a point between the inlet and outlet ports 4, 9. Whether a direct pressure relief function (FIGS. 1 and 2) or an indirect pressure relief function in the case of overheating is implemented, both configurations provide increased safety, because the pressure in the interior of the pressure vessel 1 is reduced to values below 30 bar. This ensures that even in the case of a so-called core meltdown, which may possibly be followed by melting of the pressure vessel bottom, the supporting and retaining structure of the pressure vessel 1, like the remaining nuclear reactor building structure, are only subjected, at most, to design forces. Since a core meltdown accident, for the standard pressurized-water reactor design, is an extremely improbable event, the relief valve 22 or 22' can be welded into the pipe socket 10 or the pressure vessel wall. Alternatively, a pressure-resistant flange link may be provided which permits inspection of the fusible points at certain intervals (when the pressure vessel is depressurized in any case because of fuel recharging). |
046719270 | summary | FIELD OF THE INVENTION The present invention relates to an improved nuclear fuel rod with a burnable absorber, consisting of a combination of gadolinium oxide and boron carbide particles coated with a helium impervious material, incorporated in the fissionable fuel. BACKGROUND OF THE INVENTION It is well-known that the process of nuclear fission involves the disintegration of the fissionable fuel material, usually enriched uranium dioxide, into two or more fission products of lower mass number. Among other things, the process also includes a net increase in the number of available free neutrons which are the basis for a self-sustaining reaction. When a reactor has operated over a period of time, the fuel assembly with fissionable material must ultimately be replaced due to depletion. Inasmuch as the process of replacement is time consuming and costly, it is desirable to extend the life of a given fuel assembly as long as practically feasible. For that reason, deliberate additions to the reactor fuel of parasitic neutron-capturing elements in calculated small amounts may lead to highly beneficial effects on a thermal reactor. Such neutron-capturing elements are usually designated as "burnable absorbers" if they have a high probability (or cross-secton) of absorbing neutrons while producing no new or additional neutrons or changing into new absorbers as a result of neutron absorption. During reactor operation, the burnable absorbers are progressively reduced in amount so that there is a compensation made with respect to the concomitant reduction in the fissionable material. The life of a fuel assembly may be extended by combining an initially larger amount of fissionable material as well as a calculated amount of burnable absorber. During the early stages of operation of such a fuel assembly, excessive neutrons are absorbed by the burnable absorber which undergoes transformation to elements of low neutron cross-section which do not substantially affect the reactivity of the fuel assembly in the latter period of its life when the availability of fissionable material is lower. The burnable absorber compensates for the larger amount of fissionable material during the early life of the fuel assembly, but progressively less absorber captures neutrons during the latter life of the fuel assembly, so that a long life at relatively constant fission level is assured for the fuel assembly. Accordingly, with a fuel assembly containing both fuel and burnable absorber in carefully proportioned quantity, an extended fuel assembly life can be achieved with relatively constant neutron production and reactivity. The incorporation of burnable absorber in fuel assemblies has been recognized in the nuclear industry as an effective means of increasing fuel capacity and thereby extending core life. Burnable absorbers are used either uniformly mixed with the fuel (i.e., distributed absorber) or are placed discretely as separate elements in the reactor, as separate burnable absorber rods, so arranged that they burn out or are depleted at about the same rate as the fuel. Thus, the net reactivity of the core is maintained relatively constant over the active life of the core. Among the various burnable absorbers that have been mixed with fuel as a distributed absorber, gadolinium oxide has been found to be an excellent absorber due to its extremely high thermal absorption cross-section. Enriched uranium dioxide, with a high U-235 isotope content, and gadolinium oxide, as a mixture, has thus been previously used in formation of nuclear fuel pellets. Where gadolinium oxide is used as the burnable absorber in fuel pellets, a disadvantage exists in that the absorbing quality of the gadolinium oxide is relatively stable over a period of time but then decreases very quickly and is lost. In effect, the burn-out rate of gadolinium oxide is faster than desired. Also, a problem exists where higher amounts of gadolinium oxide are added to uranium dioxide in the manufacture of fuel pellets because of physical limitations such as clumping of the material during fabrication. The use of sintered microspheres of gadolinium oxide in a nuclear fuel has been suggested in U.S. Pat. No. 3,759,786, which discusses the use of coated gadolinium oxide bodies, such as molybdenum coated gadolinium oxide, prepared in microsphere form, as consumable absorbers in uranium dioxide. Uranium borides and borides of metals having a low neutron-capturing section, such as zirconium, prepared in microsphere form are also suggested for incorporation into uranium dioxide. Boron compounds, such as boron carbide are also known for use as burnable absorbers. While boron compounds containing the isotope B.sup.10 are usable as burnable absorbers, they do not have the absorption of gadolinium oxide and thus must be provided in larger amounts, which thus must displace some fuel in pellets. Also, boron when used as a burnable absorber produces helium gas during burn-out, which gas produces undesirable pressures within the fuel element. In instances where boron compounds have been used in separate burnable poison rods, in conjunction with conventional fuel rods, coated boron carbide has been used, with a coating over the boron carbide particles to contain such helium. U.S. Pat. No. 3,356,618, for example, teaches the formation of coated boron particles where the boron particles have an inert coating, such as a carbide, nitride or carbo-nitride of a refractory metal such as zironium, formed thereon and the coated particles are dispersed in a metal matrix, such as iron, cobalt, nickel, aluminum and zirconium, for use in neutron absorption control elements for a nuclear reactor. In U.S. Pat. No. 3,855,061, which relates to a nuclear fuel plate containing microspheroidal particles of a fuel, the fuel particles may be coated with niobium, nickel, alumina, pyrolytic graphite or other materials. This patent also suggests that the microsphere route can be used to introduce nuclear poisons such as gadolinium or samarium to a fuel, or introduction of microspheres of a neutron poison, or neutron moderating materials such as boron carbide to the fuel when desired. It is an object of the present invention to provide a nuclear fuel element which combines the benefits of gadolinium oxide and boron carbide as absorbers in a single fuel pellet. It is another object of the present invention to provide a nuclear fuel which exhibits improved nuclear reactor power distribution control than is provided by the use of gadolinium oxide alone as a burnable absorber. SUMMARY OF THE INVENTION A nuclear fuel rod assembly has a fuel rod that contains a hybrid burnable absorber-nuclear fuel composition. Pellets of a fissionable material, such as uranium dioxide, contain gadolinium oxide and coated boron carbide particles. The boron carbide particles are present in the pellet in an amount of between about 0.02 to 1.0 percent by weight of the pellet and are of a particle size of between 20 to 100 microns in diameter. The coating on the boron carbide pellets is about 0.5 to 10 microns in thickness and is a helium gas-impervious coating, such as niobium, nickel, zirconium, or the like. |
047626470 | summary | BACKGROUND OF THE INVENTION This invention relates to a method for the management of low level radioactive nuclear power plant wastes, in particular the compaction and disposal of beadand powdered ion exchange resins mixed with a filter aid. One of the common low-level radioactive waste products produced by nuclear power plants is ion exchange resins. These resins are used to process water that circulates through the core of the nuclear reactor or steam generator. While ion exchange resin remove ion contaminants from plant coolant water, the filter aid removes undissolved particulates. A filter aid is any material such as cellulose layered on a filter cartridge along with the powdered resin to remove solid material. The resin and filter aid do not chemically react at the water temperature encountered in processing water from the nuclear plant, usually below about 60.degree. C. Elevated temperatures, those much above about 60.degree. C., are not usual and water of 100.degree. C. or greater is not encountered because the processing system is not pressurized. Bead type resins are usually used in pressurized water reactor type plants to remove ions, but are not mixed with a filter aid since filtration is not intended. Boiling water reactor type plants use the powered resin with cellulose filter aid for the dual purpose of ion exchange and filtration. The resins, and cellulose when spent retain residual radioactivity and have to be disposed of in a safe manner which usually requires burial in a land fill. The current practice is to encapsulate the resin in a matrix of cement or polymer to ensure adequate mechanical integrity as well as preventing leaching of radioactive substances from the resin by ground water. The disadvantage of this method is that it increases the volume of material that needs to be disposed. The price of disposal is closely related to the volume of material. Another method recently developed uses high integrity containers to hold the resins and cellulose without the addition of cement. The containers are designed to maintain boundary integrity for several hundred years. But again, the cost of transporting and burying the wastes is based upon their volume. Significant cost savings can be realized if the volumes are reduced. It is the object of this invention, therefore, to describe a method whereby the volume of ion exchange resins mixed with filter aid can be significantly reduced, and it is a further object of this invention to describe a method wherein the volume reduced resins display the ability to resist dissolution of the radioactive material in the presence of water. SUMMARY OF THE INVENTION The above-described resins are particulates having a void factor of approximately 30-40%. By applying the proper mechanical force or pressure the particles can be forced closer together, reducing the void fraction and thereby the total volume. At an elevated temperature cross-link bonds in the resin are broken and the resin does not spring back. It has been found that the volume of a mixture of 30 to 60 weight percent (w%) ion exchange resin and 40 to 70 w% cellulose filter aid can be reduced significantly by removing the water from the mixture, raising the mixture to an elevated temperature of about 230.degree. C. and compacting the heated resin with a force of at least 2000 psi. A further advantage is realized in that the resin sinters to form a monolith that is physically stable in water. |
description | This application is a National Stage (371) entry of PCT Application PCT/US2006/028890, filed Jul. 26, 2006, and published under the PCT Articles in English as WO 2007/019053 A1 on Feb. 15, 2007. PCT/US2006/028890 claimed the benefit of U.S. Provisional Application No. 60/705,376, filed Aug. 4, 2005. The present application claims the benefit of U.S. Provisional Application Ser. No. 60,705,376, filed Aug. 4, 2005, which is hereby incorporated herein by reference. This application also contains subject matter which is related to the subject matter of the following applications, each of which is hereby incorporated herein by reference in its entirety: “AN OPTICAL DEVICE FOR DIRECTING X-RAYS HAVING A PLURALITY OF OPTICAL CRYSTALS,” by Zewu Chen, U.S. Ser. No. 11/048,146, filed Feb. 1, 2005, which application is a continuation of PCT Application PCT/US2003/023412, filed Jul. 25, 2003, and published under the PCT Articles in English as WO 2004/013867 A2 on Feb. 12, 2004, which PCT application claimed priority to U.S. Provisional Application No. 60/400,809, filed Aug. 2, 2002. This invention relates in general to x-ray optics. More particularly, the present invention relates to an arrangement of x-ray optics and an associated system for producing a monochromatic x-ray beam of a highly controlled, small spot size, on a sample for x-ray analysis, including for example x-ray fluorescence and x-ray diffraction analysis. As discussed in certain commonly-assigned, co-pending, published U.S. patent applications related to wavelength dispersive x-ray fluorescence (XRF) systems, the use of monochromating optics in the excitation and/or detection paths provides various advantages, including the ability to narrow the spectrum of wavelengths present at the sample under analysis, thereby increasing the signal-to-background ratio of the system and improving analysis results (e.g., see U.S. application No. 60/336,584 filed Dec. 4, 2001, and entitled “X-Ray Tube and Method and Apparatus for Analyzing Fluid Streams Using X-Rays,” perfected as PCT Application PCT/US02/38792-WO03/048745, entitled “X-Ray Tube and Method and Apparatus for Analyzing Fluid Streams Using X-Rays;” and U.S. application No. 60/299,371 filed Jun. 19, 2001, and entitled “XRF System Including Focusing Optic on Excitation Side and Monochromatic Collection,” perfected as PCT Application PCT/US02/19272-WO02/103710, entitled “Wavelength Dispersive XRF System Using Focusing Optic for Excitation and a Focusing Monochromator for Collection” all of which are incorporated by reference herein in their entirety). Similar benefits of such monochromatization can also be applicable to other types of x-ray analysis systems, including, for example, x-ray diffraction systems depending on the particular application. Though not directly addressing monochromatization, advanced diffraction systems are addressed in, for example, commonly assigned U.S. application No. 60/492,400 filed Aug. 4, 2003 entitled “In-Situ X-Ray Diffraction System Using Sources and Detectors at Fixed Angular Positions,” perfected as PCT Application PCT/US04/25112-WO2005031329 of the same title; and application No. 60/489,047 filed Jul. 22, 2003 entitled “Method and System for X-Ray Diffraction Measurements Using an Aligned Source and Detector Rotating Around a Sample Surface,” perfected as U.S. application Ser. No. 10/893,511 filed Jul. 16, 2004, all of which are incorporated by reference herein in their entirety. In addition to monochromatic beams, small, intense x-ray beam spot sizes at the sample are also of significant interest in certain x-ray analysis systems. Small spot sizes are usually correlated with increased x-ray intensity as a function of the input x-ray source power and the optic focusing capabilities. Increased power on the sample leads to improved analysis results. However, certain applications require small spot sizes for reasons other than intensity—smaller spot sizes enable higher spatial resolution and therefore more precise elemental mapping of the sample; as well as the ability to isolate certain features on a sample while preventing interfering fluorescence/diffraction returns from other adjacent features on the sample. Excitation beam spot sizes, though dramatically improved in the last few years, are still larger than many integrated circuit (IC) feature sizes now in use (e.g., scribe lines), and those planned for the “nanosystems” of the future. Small spot sizes also enable small sample aperture sizes in, e.g., high-pressure on-line analysis systems, as discussed at length in the above-incorporated U.S. patent applications. The monochromating optics discussed in the above-incorporated U.S. patent applications (e.g., doubly curved crystals) can achieve small spot sizes, but usually at the expense of aperture size, and (as fundamentally a device which is imaging the source) as a function of the x-ray source spot size, which may be too large. What is required, therefore, are techniques, methods and systems which exploit the benefits of x-ray monochromatization in e.g., XRF systems, while also providing small, intense, x-ray beam spot sizes. The shortcomings of the prior art are overcome and additional advantages are provided through the present invention which in one aspect is an x-ray system or method for exciting a sample under x-ray analysis, using a curved monochromating optic for directing a monochromatic x-ray beam from an x-ray source towards a first focal point. A second optic is positioned within, and receives, the monochromatic x-ray beam, and directs a focused x-ray beam towards a second focal point on the sample. A detector is positioned near the sample to collect radiation from the sample as a result of the focused x-ray beam. The curved monochromating optic may have an optical surface, the optical surface being doubly-curved, e.g., a doubly curved crystal optic or a doubly curved multilayer optic. The second optic may be a polycapillary optic, or a monocapillary optic. The curved monochromating optic produces a beam spot size at the first focal point larger than a beam spot size produced by the second optic at the second focal point, therefore, a beam spot size on the sample is thereby reduced using the second optic. The second optic may be positioned within the monochromatic x-ray beam, before the first focal point, thereby receiving the monochromatic x-ray beam as it converges toward the first focal point. In this case, the second optic is a convergent-beam-to-point focusing optic. The second optic may also be positioned within the monochromatic x-ray beam, after the first focal point, thereby receiving the monochromatic x-ray beam as it diverges from the first focal point. In this case, the second optic is a divergent-beam-to-point focusing polycapillary optic. The x-ray source may be an inexpensive, electron-bombardment-type x-ray tube, having a low power of less than 100 watts. A controller may be provided for monitoring and/or controlling the position of the sample, second focal point, and/or the detector to provide an accurate indication of the location of the focal point on the sample. The disclosed embodiments of the present invention offer the advantages of monochromatization in the excitation path (which increases the signal-to-background of the system) as well as smaller spot sizes than those otherwise attainable (which provide better spatial resolution, the ability to distinguish between different types of sample features, and/or smaller sample apertures where needed). In x-ray analysis systems, high x-ray beam intensity, and small beam spot sizes, are important to reduce sample exposure times, increase spatial resolution, and consequently, improve the signal-to-background ratio and overall quality of x-ray analysis measurements. In the past, expensive and powerful x-ray sources, such as rotating anode x-ray tubes or synchrotrons, were the only options available to produce high-intensity x-ray beams. Recently, the development of x-ray optic devices has made it possible to collect the diverging radiation from an x-ray source by focusing the x-rays. A combination of x-ray focusing optics and small, low-power x-ray sources can produce x-ray beams with intensities comparable to those achieved with more expensive devices. As a result, systems based on a combination of small, inexpensive x-ray sources, excitation optics, and collection optics have greatly expanded the availability and capabilities of x-ray analysis equipment in, for example, small laboratories or at line or online production or processing applications. Monochromatization of x-ray beams in the excitation and/or detection paths is also useful, as discussed above. One existing x-ray monochromatization technology is based on diffraction of x-rays on optical crystals, for example, germanium (Ge) or silicon (Si) crystals. Curved crystals can provide deflection of diverging radiation from an x-ray source onto a target, as well as providing monochromatization of photons reaching the target. Two common types of curved crystals are known as singly-curved crystals and doubly-curved crystals (DCC). Using what is known in the art as Rowland circle geometry, singly-curved crystals provide focusing in two dimensions, leaving x-ray radiation unfocused in the third or orthogonal plane. Doubly-curved crystals provide focusing of x-rays from the source to a point target in all three dimensions. This three-dimensional focusing is referred to in the art as “point-to-point” focusing. The point-to-point focusing property of doubly-curved crystals has many important applications in, for example, material science structural analysis. Depending on the bending radii of the doubly-curved crystal in the Rowland optic circle plane, curved crystals further divide into Johansson and Johann types. Typically, Johansson geometry requires crystals to have a curvature that is equal to the radius of the Rowland circle, while Johann geometry configuration may have a curvature twice the radius of the Rowland circle. The present invention, in one embodiment, is directed to an x-ray analysis system using monochromatic excitation to improve signal-to-background ratio significantly, compared to the use of polychromatic excitation. As discussed above, doubly-curved crystal (DCC) optics can provide high intensity monochromatic focused beam using a low power x-ray tube. However, since the focal spot of a DCC optic is strongly dependent on the x-ray tube spot size and also the optic aperture, it can be difficult to achieve a beam spot of less than the spot size of the source (in one source example, 70 μm). The present invention is directed to a novel x-ray optic which enables small spot, monochromatic beams for mapping of elements, with a high spatial resolution. For example, a spot size of less than 17 μm can be achieved (compared to an imaged source having a 70 μm spot). A spot size of about 1 micrometer can be achieved with a monocapillary optic. Two beam energies, Mo Ka and Cu Ka, have been considered for this approach. With reference to FIG. 1, x-ray analysis system 10 includes a source 12, and a curved, monochromating optic 14. In general (though not required), optic 14 images an x-ray beam produced by source 12 towards an output focal point 15 (F1). A second, focusing optic 16 is placed in this beam path (which is converging on point 15), and receives the beam prior to its focal point 15 at the optic's input end. The second optic is designed, e.g., as a convergent-beam-to-point optic, and provides a second level of focusing to its own output focal point 17 (F2) on a sample 11 under analysis. (Note, focal point 15 is discussed only as a basis to describe the operation of optic 14. Since the beam is re-directed prior to this point, x-rays may not actually “exist” at that point.) Convergent-beam-to-point optics can be implemented using polycapillary optics such as those disclosed in commonly assigned, X-Ray Optical Systems, Inc. U.S. Pat. Nos. 5,192,869; 5,175,755; 5,497,008; 5,745,547; 5,570,408; and 5,604,353; the entirety of which are incorporated herein by reference herein in their entirety. In addition to the “barrel” and “half-barrel” optics disclosed in those patents, polycapillary optics such as a conical optic, with non-bent but contracting capillary diameters can also be used. Monocapillary optics can also be used, such as those disclosed in commonly assigned, X-Ray Optical Systems, Inc. U.S. Pat. No. 5,747,821; the entirety of which is incorporated herein by reference herein in its entirety. In general, such optics should be capable of re-focusing x-ray energy, and be of the reflecting (rather than absorbing) type. Such focusing optics can capture x-ray beams and focus to small beam spot sizes on the order of 17 μm or less, depending on the input energy level. This beam spot size is smaller than that attainable using the source 12/optic 14 configuration alone, especially considering, as discussed above, that optic 14 may be limited to imaging the spot size of the source 12, which in one example is about 70 μm (using an x-ray tube available from Oxford Instruments—#5011). In the disclosed configuration, the source 12/optic 14 combination provides a “virtual source” for focusing optic 16, which in turn provides the additional level of focusing. In an alternate embodiment of the present invention, with reference to FIG. 2, x-ray analysis system 30 includes a source 32, and a curved, monochromating optic 34. In general (though not required), optic 34 images an x-ray beam produced by source 32 towards an output focal point 35 (F1). As shown, and consistent with the operation of point imaging optics, this beam may first converge towards its focal point 35, and then diverges beyond this focal point. In this embodiment, a second, focusing optic 36 is placed in this beam path (which is diverging beyond focal point 35), and receives this beam beyond its focal point at the optic's input end. The second optic is designed as a divergent-beam-to-point optic, and provides a second level of focusing to its own output focal point 37 (F2) on a sample under analysis 31. Divergent-beam-to-point optics can be implemented using polycapillary optics such as those disclosed in the above-incorporated U.S. Pat. Nos. 5,192,869; 5,175,755; 5,497,008; 5,745,547; 5,570,408; and 5,604,353; polycapillary conical optics; and monocapillary optics as discussed above. Again, such optics should be capable of re-focusing x-ray energy, and be of the reflecting (rather than absorbing) type. As in the first embodiment, such optics can capture x-ray beams and focus to small beam spot sizes on the order of 17 μm or less, depending on the input energy level. This beam spot size is smaller than that attainable using the source 32/optic 34 configuration alone, especially considering, as discussed above, that optic 34 may be limited to imaging the spot size of the source 32, which in one example is about 70 μm (using an x-ray tube available from Oxford Instruments—#5011). In this configuration, the source 32/optic 34 combination provides a “virtual source” for focusing optic 36, which in turn provides an additional level of focusing. Also shown in FIGS. 1 and 2 are portions of a shielding or cabinet 20, 40, respectively. This shielding or cabinet can include a relatively small aperture for passing the x-ray beam from DCC 14, 34, respectively, at a point on the beam where it has been significantly narrowed by the optic 14, 34. This shielding can be advantageously used to isolate the detection subsystem from x-ray background noise generated by the source and first optic. Eliminating this background noise can dramatically improve measurement results. Though point focusing is disclosed in these embodiments, those skilled in the art will recognize that the present invention can extend to fine-line focusing as well, in which case the monochromating optic focuses to a first focal line (or rectangle), and a polycapillary optic (e.g., in a generally rectangular shape) re-focuses the line to a finer line. Or, in another embodiment, the monochromating may produce a parallel beam, of, for example, substantially uniform cross-sectional shape along its beam path. The second, focusing optic can be a parallel-to-point focusing optic placed anywhere along this path (and in fact translatable along this path) to provide a variable output focal spot position. This can assist in elemental mapping of a sample, using a small, but moveable beam spot on the sample, using only the movement of the second optic. Also, in either of the embodiments of FIGS. 1 and 2 above (especially for diffraction) the second optic could be either a convergent-beam-to parallel optic, or (preferred) divergent-beam-to-parallel optic; otherwise known as a collimating optic. These types of collimating optics are also disclosed in the above-incorporated U.S. patent applications. In all of these cases, it is important to note that both optics of the present invention are changing the convergence/divergence path of the beam in some appreciable way. This beam convergence/divergence shaping is distinguished from other flat crystal optics, or slit optics. Therefore, for the purposes of this application, the terms “focus,” “focusing” or “focused” connote some appreciable change of the convergence or divergence of the input beam, toward a focal “area” (e.g., line, point, rectangle, etc.) Point focusing optics are shown in FIGS. 1 and 2 for illustration only and those skilled in the art will recognize that the focal “areas” of the optics can be customized to any particular shape. In both embodiments, detectors 18 and 38 respectively are arranged over/about the sample to collect the resultant x-ray radiation from the sample (e.g., diffraction, fluorescence, etc.) as discussed in the above-incorporated U.S. patents and patent applications, and provide the requisite analytical results through systems 19/39. In fact, in certain applications detection path optics can be used for additional monochromatization (in, for example, the above-incorporated XRF patent applications which discuss wavelength dispersive spectrometry and the relevant detectors) and angular filtering (in, for example, the above-incorporated XRD patent applications which discuss a variety of applicable detection schemes). All of the components, including sample positioning stages (not shown), can be controlled and/or monitored by a controller 19/39, which can collect positional data from any of the elements/beams shown, and provide trace element mapping across the sample according to the resolution offered by the small x-ray beam spot size of the present invention, i.e., 17 μm or less. This is a fundamental advantage of the present invention—the ability to perform trace element mapping over a sample surface with a high degree of accuracy driven by the small beam spot size. CCD cameras can be used within these systems to obtain the positional data required. Example DCC Optic Technology: The curved, monochromating optics 14 and 34 can comprise various optical devices, including a doubly-curved crystal (DCC) optic or a doubly-curved multilayer optic. One embodiment of such a doubly-curved optical device is depicted in FIGS. 3 and 3A, and is described in detail in U.S. Pat. No. 6,285,506 B1, issued Sep. 4, 2001, the entirety of which is hereby incorporated herein by reference. In the embodiment of FIG. 3, a doubly-curved optical device includes a flexible layer 110, a thick epoxy layer 112 and a backing plate 114. The structure of the device is shown further in the cross-sectional elevational view in FIG. 3A. In this device, the epoxy layer 112 holds and constrains the flexible layer 110 to a selected geometry having a curvature. Preferably, the thickness of the epoxy layer is greater than 20 μm and the thickness of the flexible layer is greater than 5 μm. Further, the thickness of the epoxy layer is typically thicker than the thickness of the flexible layer. The flexible layer can be one of a large variety of materials, including: mica, Si, Ge, quartz, plastic, glass etc. The epoxy layer 112 can be a paste type with viscosity in the order of 103 to 104 poise and 30 to 60 minutes pot life. The backing plate 114 can be a solid object that bonds well with the epoxy. The surface 118 of the backing plate can be flat (FIG. 3A) or curved, and its exact shape and surface finish are not critical to the shape and surface finish of the flexible layer. In the device of FIGS. 3 & 3A, a specially prepared backing plate is not required. Surrounding the flexible layer may be a thin sheet of protection material 116, such as a thin plastic, which is used around the flexible layer edge (see FIG. 3A). The protection material protects the fabrication mold so that the mold is reusable, and would not be necessary for a mold that is the exact size or smaller than the flexible layer, or for a sacrificial mold. Doubly-curved optical devices, such as doubly-curved crystal (DCC) optics, are now used in material analysis to collect and focus x-rays from a large solid angle and increase the usable flux from an x-ray source. Three-dimensional focusing of characteristic x-rays can be achieved by diffraction from a toroidal crystal used with a small x-ray source. This point-to-point Johan geometry is illustrated in FIG. 4. The diffracting planes of each crystal optic element 200 can be parallel to the crystal surface. If the focal circle 210 containing a point source and the focal point has radius R0, then the crystal surface has, for example, a radius R of curvature of 2R0 in the plane of the focal circle and a radius of curvature of r=2R0 sin 2θBrag in the perpendicular plane, with the radius centered on a line segment drawn between the source and the focal point. X-rays diverging from the source, and incident on the crystal surface at angles within the rocking curve of the crystal will be reflected efficiently to the focal or image point. The monochromatic flux density at the focal point for a DCC-based system is several orders of magnitude greater than that of conventional systems with higher power sources and similar source to object distances. This increase yields a very high sensitivity for use in many different applications, including (as described herein) x-ray fluorescence and diffraction. As a further enhancement, FIG. 4 illustrates that the optical device may comprise multiple doubly-curved crystal optic elements 200 arranged in a grid pattern about the Rowland circle. Such a structure may be arranged to optimize the capture and redirection of divergent radiation via Bragg diffraction. In one aspect, a plurality of optic crystals having varying atomic diffraction plane orientations can be used to capture and focus divergent x-rays towards a focal point. In another aspect, a two or three dimensional matrix of crystals can be positioned relative to an x-ray source to capture and focus divergent x-rays in three dimensions. Further details of such a structure are presented in the above-incorporated, co-pending U.S. patent application Ser. No. 11/048,146, entitled “An Optical Device for Directing X-Rays Having a Plurality of Optical Crystals”. Example Source/Optic Technology: As discussed above, the ability to provide an efficient and economical analysis capability depends to a large extent upon source/optic technology. In that regard, certain source and optic technology can be combined for such systems, as discussed below with respect to FIG. 5 in which the basic elements of a typical compact, electron-bombardment x-ray source 300 are shown. Electron gun/filament 310 is heated (by applying a voltage) to a temperature such that electrons 312 are thermally emitted. These emitted electrons are accelerated by an electric potential difference to anode 314, which is covered with target material, where they strike within a given surface area of the anode, called the spot 318. Divergent x-rays 320 are emitted from the anode as a result of the collision between the accelerated electrons and the atoms of the target. One example of this type of x-ray tube is available from Oxford Instruments—model #5011, which operates at less than 100 watts (i.e., 75 watts) at a cost of less than $1500 per tube, in contrast to higher-power, laboratory sources which can cost many thousands, or hundreds of thousands of dollars—which is cost prohibitive for many applications. Source/optic combinations are also useable for the source 12, 32/optic 14, 34 combinations above, such as those disclosed in X-Ray Optical Systems, Inc.'s U.S. Pat. No. 5,570,408, issued Oct. 29, 1996, as well as in U.S. Provisional Application Ser. Nos. (1) 60/398,968 (filed Jul. 26, 2002, entitled “Method and Device for Cooling and Electrically-Insulating a High-Voltage, Heat-Generating Component,” and perfected as PCT Application PCT/US02/38803); (2) 60/398,965 (filed Jul. 26, 2002, entitled “X-Ray Source Assembly Having Enhanced Output Stability,” and perfected as PCT Application PCT/US02/38493); (3) 60/492,353 (filed Aug. 4, 2003, entitled “X-Ray Source Assembly Having Enhanced Output Stability Using Tube Power Adjustments and Remote Calibration”); and (4) 60/336,584 (filed Dec. 4, 2001, and entitled “X-Ray Tube and Method and Apparatus for Analyzing Fluid Streams Using X-Rays,” perfected as PCT Application PCT/US02/38792-WO03/048745, entitled “X-Ray Tube and Method and Apparatus for Analyzing Fluid Streams Using X-Rays”)—all of which are incorporated by reference herein in their entirety. The disclosed embodiments of the present invention offer the advantages of monochromatization in the excitation path (which increases the signal-to-background of the system) as well as smaller spot sizes than those otherwise attainable (which provide better spatial resolution, the ability to distinguish between different types of sample features, and/or smaller sample apertures where needed). FIGS. 6-8 show results of the present invention, when arranged according to the principles of FIG. 1. FIG. 6 shows a knife-edge scan of the resultant spot size (i.e., 19 μm) using a source with an Mo target and MoKa energy. FIG. 7 shows a knife-edge scan of the resultant spot size (i.e., 26 μm) using a source with a Cu target and CuKa energy. FIG. 8 shows an x-ray fluorescence, trace element mapping of a bone section for 4 different elements—using the small spot size produced by the present invention. Although preferred embodiments have been depicted and described in detail herein, it will be apparent to those skilled in the relevant art that various modifications, additions, substitutions and the like can be made without departing from the spirit of the invention and these are therefore considered to be within the scope of the invention as defined in the following claims. |
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048184730 | abstract | A fuel bundle in which the fuel rods are supported by top and bottom plates and held together by axially spaced straps. Several of the fuel rods are formed as tie rods. Each tie rod has a flat key at the bottom which is passed through a slot in the bottom plate. By turning the tie rod, the key is rotated to a locking position in which its axial movement is prevented. At the top, each tie rod has a flattened plug which is threaded on its rounded sides to receive a nut that has a crimping lip. The plug extends through a slot in the top plate and is threaded to the nut and the crimping lip is crimped to the plug. The tip of the plug is flat. During reconstitution of the fuel bundle, the crimp on each tie rod is broken and the nut is unscrewed while the tip of the plug is held. The top plate is then removed and each tie rod is then turned so that the key can be passed through the slot in the bottom plate. The tie rods may thus be removed and the bundle disassembled without access to the bottom plate. Each tie rod is provided with tabs to prevent axial displacement of the grids. |
description | This application claims priority under 35 U.S.C. § 120 to, and is a divisional of, U.S. patent application Ser. No. 12/272,215, filed Nov. 17, 2008, and entitled “Reactor Vessel Reflector With Integrated Flow-Through,” the contents of which are incorporated herein by reference in their entirety. The invention relates to the field of nuclear power generation, including systems designed to cool a reactor core. In nuclear reactors designed with passive operating systems, the laws of physics are employed to ensure that safe operation of the nuclear reactor is maintained during normal operation or even in an emergency condition without operator intervention or supervision, at least for some predefined period of time. A nuclear reactor 5 includes a reactor core 6 surrounded by a reactor vessel 2. Water 10 in the reactor vessel 2 surrounds the reactor core 6. The reactor core 6 is further located in a shroud 122 which surround the reactor core 6 about its sides. When the water 10 is heated by the reactor core 6 as a result of fission events, the water 10 is directed from the shroud 122 and out of a riser 124. This results in further water 10 being drawn into and heated by the reactor core 6 which draws yet more water 10 into the shroud 122. The water 10 that emerges from the riser 124 is cooled down and directed towards the annulus 123 and then returns to the bottom of the reactor vessel 2 through natural circulation. Pressurized steam 11 is produced in the reactor vessel 2 as the water 10 is heated. A heat exchanger 135 circulates feedwater and steam in a secondary cooling system 130 in order to generate electricity with a turbine 132 and generator 134. The feedwater passes through the heat exchanger 135 and becomes super heated steam. The secondary cooling system 130 includes a condenser 136 and feedwater pump 138. The steam and feedwater in the secondary cooling system 130 are isolated from the water 10 in the reactor vessel 2, such that they are not allowed to mix or come into direct contact with each other. The reactor vessel 2 is surrounded by a containment vessel 4. The containment vessel 4 is designed so that water or steam from the reactor vessel 2 is not allowed to escape into the surrounding environment. A steam valve 8 is provided to vent steam 11 from the reactor vessel 2 into an upper half 14 of the containment vessel 4. A submerged blowdown valve 18 is provided to release the water 10 into suppression pool 12 containing sub-cooled water. Water 10 circulates through the reactor vessel 2 as a result of temperature and pressure differentials that develop as a result of heat generation through reactor operation and through heat exchange with the secondary cooling system 130. Accordingly, the efficiency of the circulation depends on the relative local thermal properties of the water 10 in the reactor module 5 as well as its physical design and geometry. Reactor core reflectors are provided to improve a performance of the reactor core and associated fission events. As a result of the close proximity of the reflector to the reactor core, the reflector tends to heat up. Primary coolant is used to cool the reflector, and coolant circulation through the reactor core is reduced as a result of the supply of water 10 needed for cooling the reactor core reflector. Conventional nuclear reactors must therefore rely on increased coolant volume, pumps, or other redundant system components to ensure sufficient performance. A power module is disclosed herein, as comprising a reactor core and a reflector partially surrounding the reactor core to improve a neutron efficiency of the power module. The reflector comprises one or more inlets located adjacent the reactor core, wherein the one or more inlets are configured to receive coolant that has passed through at least a portion of the reactor core prior to entering the one or more inlets. A nuclear reactor module is disclosed herein, as comprising a reactor core and a reactor housing that surrounds the reactor core about its sides, wherein the reactor housing is configured to direct coolant through the reactor core. A neutron reflector is located between the reactor core and the reactor housing, wherein the neutron reflector comprises a plurality of inlet ports facing the reactor core. The neutron reflector further comprises a plurality of outlet ports fluidly connected to the inlet ports to direct a portion of the coolant through the neutron reflector. A method is disclosed herein, as comprising circulating a primary coolant into a reactor housing, and partitioning the primary coolant into a first portion and a second portion. The first portion passes entirely through a reactor core located in the reactor housing, and the second portion passes partially through the reactor core before entering an inlet of a neutron reflector. The second portion of the primary coolant is directed through the neutron reflector. The invention will become more readily apparent from the following detailed description of example embodiments of the invention, which proceeds with reference to the accompanying drawings. Various embodiments disclosed or referred to herein may be operated consistent, or in conjunction, with features found in co-pending U.S. application Ser. No. 11/941,024 which is herein incorporated by reference in its entirety. FIG. 2 illustrates a power module assembly 25 comprising an internally dry containment vessel 54. The containment vessel 54 is cylindrical in shape, and has ellipsoidal, domed or hemispherical upper and lower ends. The entire power module assembly 25 may be submerged in a pool of water 16 which serves as an effective heat sink. The containment vessel 54 may be welded or otherwise sealed to the environment, such that liquids and gas do not escape from, or enter, the power module assembly 25. The containment vessel 54 may be supported at any external surface. In one embodiment, the containment vessel 54 is suspended in the pool of water 16 by one or more mounting connections 180. The pool of water 16 and the containment vessel 54 may further be located below ground 9 in a reactor bay 7. A reactor vessel 52 is located or mounted inside the containment vessel 54. An inner surface of the reactor vessel 52 may be exposed to a wet environment including a coolant 100 or liquid, such as water, and an outer surface may be exposed to a dry environment such as air. The reactor vessel 52 may be made of stainless steel or carbon steel, may include cladding, and may be supported within the containment vessel 54. The power module assembly 25 may be sized so that it can be transported on a rail car. For example, the containment vessel 54 may be constructed to be approximately 4.3 meters in diameter and approximately 17.7 meters in height (length). Refueling of the reactor core 6 may be performed by transporting the entire power module assembly 50 by rail car or overseas, for example, and replacing it with a new or refurbished power module assembly which has a fresh supply of fuel rods. The containment vessel 54 encapsulates and, in some conditions, cools the reactor core 6. It is relatively small, has a high strength and may be capable of withstanding six or seven times the pressure of conventional containment designs in part due to its smaller overall dimensions. Given a break in the primary cooling system of the power module assembly 25 no fission products are released into the environment. Decay heat may be removed from the power module assembly 25 under emergency conditions. The reactor core 6 is illustrated as being submerged or immersed in a primary coolant 100, such as water. The reactor vessel 52 houses the coolant 100 and the reactor core 6. A reactor housing 20 comprises a shroud 22 in a lower portion and a riser 24 in an upper portion of the reactor housing 20. The riser 24 may be substantially cylindrical in shape. The shroud 22 surrounds the reactor core 6 about its sides and serves to direct the coolant 100 (shown as coolant flow 26, 28) up through the center of the riser 24 located in the upper half of the reactor vessel 52, then back down the annulus 23, as a result of natural circulation of the coolant 100. In one embodiment, the reactor vessel 52 is approximately 2.7 meters in diameter and includes an overall height (length) of approximately 13.7 meters. The reactor vessel 52 may include a predominately cylindrical shape with ellipsoidal, domed or hemispherical upper and lower ends. The reactor vessel 52 is normally at operating pressure and temperature. The containment vessel 54 is internally dry and may operate at atmospheric pressure with wall temperatures at or near the temperature of the pool of water 16. The containment vessel 54 substantially surrounds the reactor vessel 52 and may provide a dry, voided, or gaseous environment identified as containment region 44. Containment region 44 may comprise an amount of air or other fill gas such as Argonne or other noble gas. The containment vessel 54 includes an inner surface or inner wall which is adjacent to the containment region 44. The containment region 44 may include a gas or gases instead of or in addition to air. In one embodiment, the containment region 44 is maintained at or below atmospheric pressure, for example as a partial vacuum. Gas or gasses in the containment vessel may be removed such that the reactor vessel 52 is located in a complete or partial vacuum in the containment region 44. During normal operation, thermal energy from the fission events in the reactor core 6 causes the coolant 100 to heat. As the coolant 100 heats up, it becomes less dense and tends to rise up through the riser 24. As the coolant 100 temperature reduces, it becomes relatively denser than the heated coolant and is circulated around the outside of the annulus 23, down to the bottom of the reactor vessel 52 and up through the shroud 22 to once again be heated by the reactor core 6. This natural circulation causes the coolant 100 (shown as coolant flow 26, 28) to cycle through the heat exchanger 135 of FIG. 1, transferring heat to a secondary coolant, such as the secondary cooling system 130 of FIG. 1, to generate electricity. FIG. 3 illustrates a cross sectional side view of an embodiment of a power module assembly 30 comprising a reactor core 6 surrounded by a neutron reflector 35. The neutron reflector 35 may partially or completely surround the reactor core 6 about its sides. In one embodiment, the neutron reflector 35 is externally cylindrical in shape, and internally matched to the reactor core perimeter. The neutron reflector 35 may be manufactured using stainless steel. The power module assembly 30 is illustrated as including a spray and one or more heaters to help control pressure within the reactor vessel 52. The neutron reflector 35 may be located between the reactor housing 22 and the reactor core 6. In one embodiment, the neutron reflector 35 is integrated with the reactor housing 22, such that an outer surface of the neutron reflector 35 forms a part of the downcomer region. Uranium or other reactor core materials may fission by splitting into smaller nuclei. The fission event also results in a release of a few neutrons and a large release of energy in the form of fission product motion, gamma rays (gammas), neutrons, and neutrinos. The neutron reflector 35 maintains a neutron efficiency of the power module assembly 30, wherein neutrons that escape the reactor core 6 during fission events are reflected back towards the reactor core 6. The neutron reflector 35 also absorbs or moderates fission products (e.g. gammas and neutrons) to protect the reactor vessel 52 from damage. As a result of absorbing some of the gammas and neutrons, the neutron reflector 35 tends to heat up over time. Primary coolant 28 flows through the reactor core 6 to become relatively hot coolant T H. Coolant flow 26 exiting the top of the riser 24 is directed down the annulus, after which heat from the hot coolant This transferred to a heat exchanger, and then recirculated back down the reactor vessel 52 to the reactor core 6. The coolant flow 28 that circulates up through the reactor core 6 also operates to cool a surface of the neutron reflector 35, but as the thickness of the neutron reflector 35 is increased to improve neutron efficiencies, this cooling of the surface may be insufficient to protect the neutron reflector 35 from overheating. FIG. 4 illustrates a partial view of a power module assembly comprising a reactor core 6 and a conventional reflector 45. Coolant flow Tc circulating past the reflector 45 is partially diverted to flow through a vertical bypass hole 40 provided through the length of the reflector 45. The remainder of the coolant flow Tc flows through the reactor core 6. Accordingly, coolant flow Tc is divided into two parts, including coolant flow 46 which passes through the reactor core 6, and coolant flow 47 which passes through the bypass holes 40 of reflector 45. Either of coolant flows 46, 47 (taken individually) include smaller coolant flow rates as compared with coolant flow Tc. Coolant flow 47 does not pass through the reactor core 6. Accordingly, the effective flow rate of coolant flow Tc that passes through the reactor core 6 is reduced by an amount equal to coolant flow 47. This reduces an operating efficiency of the power module, as additional coolant must be provided to the system, or a flow rate of the coolant must be augmented by pumps to make up for the coolant flow 47 that is diverted through the reflector 45. FIG. 5 illustrates a partial view of an example power module assembly comprising a reactor core 6 and a novel neutron reflector 50. Reactor housing 22 surrounds the reactor core 6 about its sides to direct the coolant Tc through the reactor core 6. Neutron reflector 50 is located between the reactor housing 22 and the reactor core 6. In one embodiment, the neutron reflector 50 partially or completely surrounds the reactor core 6 about its sides to improve a neutron efficiency of the power module. The neutron reflector 50 comprises one or more inlets IN1, IN2 located adjacent the reactor core 6, wherein the one or more inlets IN1, IN2 are configured to receive coolant F1 that has passed through at least a portion of the reactor core 6 prior to entering the one or more inlets IN1, IN2. The neutron reflector 50 comprises a lower end 51, an upper end 55, and a side wall 53 that faces the reactor core 6, wherein the one or more inlets IN1, IN2 are located in the side wall 53. The first inlet IN1 is located at a lower elevation than the second inlet IN2. The first inlet IN1 is configured to receive coolant that passes through a lower portion of the reactor core 6. The second inlet IN2 is located at a higher elevation than the first inlet IN1. The second inlet IN2 is configured to receive coolant that passes through an upper portion of the reactor core 6. By drawing the coolant into the inlets IN1, IN2 at the different elevations rather than at the bottom of the neutron reflector 50, all the coolant is drawn, at least partially, through the heated reactor core 6, rather than bypassing the heated reactor core 6. By drawing the coolant through inlets located at different heights, the coolant may immediately be used to cool down different regions of the neutron reflector 50 simultaneously, without having to wait for coolant to pass through the entire length of the neutron reflector 50. Conventional reflectors, on the other hand, preferentially cool a lower portion of the reflector, at a point where the coolant enters the bypass hole (reference FIG. 4). However, the lower portion of the reflector may in fact not be experiencing the greatest amount of heat generation due to the axial power distribution of the reactor core. Coolant Tc that circulates within the reactor vessel 52 enters the reactor core 6 as coolant flow Fo. In one embodiment, the flow rate associated with coolant T c is the same as the flow rate associated with coolant flow Fo that enters the bottom of the reactor core 6. Coolant that has partially passed through the reactor core 6 enters the one or more inlets IN1, IN2 as coolant cross-flow F1. Coolant cross-flow F1 cools down the neutron reflector 50 as it travels between the one or more inlets IN1, IN2 and one or more outlets OUT. The one or more outlets OUT are located at the top surface or upper end 55 of the neutron reflector 50. Coolant flow F2 exits the neutron reflector 6 via the one or more outlets OUT. In one embodiment, the flow rate associated with coolant flow F2 is the same as the flow rate associated with coolant cross-flow F1. Coolant flow F2 joins the coolant flow TH which exits the reactor core 6. In one embodiment, the flow rate associated with coolant flow Fo is the same as the sum of coolant flows F2 and TH. In one embodiment, coolant cross-flow F1 enters the one or more inlets IN1, IN2 as single-phase liquid. Coolant flow F2 exits the one or more outlets OUT as two-phase steam/liquid or single-phase steam. Coolant flow F2 that exits the one or more outlets OUT introduces voiding into the riser section 24 and reduces the effective coolant density within the core elevations, enhancing coolant flow rates in the power module. Increasing the flow rate also results in a smaller change in temperature across the reactor core 6. The voiding enhances flow through the reactor core 6 and also increases primary side heat transfer coefficients inside the steam generator region. Having single-phase or two-phase coolant including steam in the neutron reflector 50 reduces the amount of moderation of neutrons that otherwise occurs within the neutron reflector 50, increasing neutron reflection efficiency. As steam or boiling heat transfer allows for more efficient heat removal as compared to single phase liquid coolant, less reflector material may be removed to provide for cooling of the internal portion of the neutron reflector 50, thereby enhancing neutron reflection efficiencies, or neutron economy. Neutron efficiencies are further improved as a result of providing steam in the single-phase or two-phase coolant within the neutron reflector 50. As steam is a poor moderator of neutrons as compared to liquid coolant, fewer neutrons are moderated as a result of drawing the super heated single-phase coolant from the core region. Providing a more efficient neutron reflector results in increased flow rates through the reactor core 6, and allows the power module to be operated at an increased operating power. FIG. 6 illustrates a partial cutaway view of an example neutron reflector 60 comprising horizontal and vertical coolant flow-through channels 69. The neutron reflector 60 may be understood as being located adjacent reactor core 6 (FIG. 5). In one embodiment, the neutron reflector 60 is located between the reactor core 6 and the reactor housing 22 of FIG. 5. The neutron reflector 60 comprises a plurality of inlet ports 62 facing the reactor core 6. The neutron reflector 60 further comprises a plurality of outlet ports 64 fluidly connected to the inlet ports 62, to direct a portion of the coolant F1 through the neutron reflector 60. The inlet ports 62 are shown located at different elevations along the side wall 63. The inlet ports 62 are located on a side wall 63 of the neutron reflector 60, whereas outlet ports 64 are located on the upper end 65 of the neutron reflector 60. Coolant exits the neutron reflector 60 via the outlet ports 64 as coolant flow F2. In one embodiment, coolant flow F2 is equal to coolant cross-flow F1. Inlet ports 62 and outlet ports 64 are fluidly connected by a plurality of flow-through channels 69. The first part of flow-through channel 69 comprises an approximately horizontal section 68, whereas a second part of flow-through channel 69 comprises an approximately vertical section 67. Horizontal channel 68 is shown connected to the inlet port 62, whereas vertical channel 67 is shown connected to the outlet port 64. In one embodiment, an optional, approximately vertical channel 66 is provided to help cool down the neutron reflector 60. The optional, approximately vertical channel 66 draws coolant from the lower end of the neutron reflector 60. Spacing between inlet ports 62 may vary. For example, a first spacing Hi between inlet ports may be greater than a second spacing Hn between inlet ports. Spacing between inlet ports 62 at or near the center of the reactor core 6 may be less than spacing between inlet ports 62 at or near either end of the reactor core 6. The number or concentration of inlet ports 62 may be greater at or near the center of the reactor core 6 as compared to the number or concentration of inlet ports 62 at or near either end of the reactor core 6. In one embodiment, the spacing of the inlet ports is made to vary according to the axial power generation profile of the reactor core 6 as well as the fluid conditions. In one embodiment, the inlet port spacing is arranged to match the point at which boiling occurs within the neutron reflector 60. Multiple input ports located at different vertical positions along the reactor core 6 provide for cooling of the neutron reflector 60 in a staged manner. The channel diameter or channel size may vary to allow more or less coolant F1 to pass through anyone of the inlet ports 62. Channel size near the center of the reactor core 6 may be greater than the channel size at either end of the reactor core 6. FIG. 7 illustrates a partial cutaway view of an example neutron reflector 70 comprising a number of sloped or diagonal coolant flow-through channels 79. The sloped coolant flow-through channels 79 fluidly couple one or more inlets 72 located in a side wall 73 of the neutron reflector 70 to one or more outlets 74 located in an upper end 75 of the neutron reflector 70. Coolant cross-flow F1 that has partially passed through reactor 6 (FIG. 5) enters the one or more inlets 72, passes through the sloped coolant flow-through channels 79, and exits the outlets F2 as coolant flow F2. In one embodiment, coolant flow F2 equals coolant cross-flow F1. Inlet port spacing Hi, Hn may vary with elevation of the inlet ports 72. A height or width of the sloped coolant flow-through channels 79 may also vary, allowing more or less coolant cross-flow F1 to pass through different portions of the neutron reflector 70. In one embodiment, one or more optional, approximately vertical channels 76, 78 are provided to allow additional coolant to pass through a lower portion of the neutron reflector 70. The one ore more approximately vertical channels 76, 78 are shown connect to one of the sloped coolant flow-through channels 79, however they may alternatively connect to their own respective outlets (no shown) located in the upper end 75 of the neutron reflector. When one or more of the optional, approximately vertical channels 76, 68 are provided in the neutron reflector 70, coolant cross-flow F1 may be less than coolant flow F2. FIG. 8 illustrates a partial cutaway view of an example neutron reflector 80 comprising diagonal and vertical coolant flow-through channels 89. The coolant flow-through channels 89 fluidly couple one or more inlets 82 located in a side wall 83 of the neutron reflector 80 to one or more outlets 84 located in an upper end 85 of the neutron reflector 80. Coolant cross-flow F1 that has partially passed through reactor core 6 (FIG. 5) enters the one or more inlets 82, passes through the sloped coolant flow-through channels 89, and exits the outlets F2 as coolant flow F2. In one embodiment, coolant flow F2 equals coolant cross-flow F1. The first part of channel 89 comprises and angled or sloped section 88, whereas a second part of channel 89 comprises an approximately vertical section 87. Horizontal channel 88 is shown connected to the inlet port 82, whereas the approximately vertical channel 87 is shown connected to the outlet port 84. In one embodiment, none of the channels or ports are connected to the bottom end 81 of the neutron reflector 80. The various embodiments of the neutron reflector may comprise any number or combination of horizontal, vertical, sloped, diagonal or otherwise oriented channels, and those embodiments illustrated herein are not intended to limit any such combinations. Furthermore, the number of inlet ports may not equal the number of outlet ports, as one or more channels may connect multiple ports together. FIG. 9 illustrates an elevated perspective view of an example neutron reflector 90 comprising a plurality of layered plates 92, 94, 96. FIG. 9 may be understood as a method of manufacturing or a method of assembling a neutron reflector, wherein FIG. 9 illustrates an exploded view of the neutron reflector 90. When assembled, lower plate 92 is adjacent to and in contact with plate 94, whereas plate 94 is adjacent to and in contact with upper plate 96. An upper surface of upper plate 96 may be understood as providing an upper end 95 of the neutron reflector 90. A lower surface of lower plate 92 may be understood as providing a lower end 91 of the neutron reflector 90. A side surface of one or all of the plates 92, 94, 96 may form the side wall 93 of the neutron reflector 90. The neutron reflector 90 comprises the plurality of plates 92, 94, 96 layered together, wherein a flow-through channel 99 is formed, at least in part, between adjacent plates 92, 94. Flow-through channel 99 fluidly connects at least one of the inlet ports INLET to at least one of the outlet ports OUTLET. The flow-through channel 99 comprises an approximately horizontal channel 98 that passes between the adjacent plates 92, 94. An upper portion of the approximately horizontal channel 98 is recessed into a lower surface of a first plate, such as middle plate 94, wherein the approximately horizontal channel 98 comprises a lower portion that is bounded by an upper surface of a second plate, such as lower plate 92. The lower surface of the middle plate 94 is located adjacent to the upper surface of the lower plate 92 when the neutron reflector 90 is fully assembled. Another channel is shown formed between middle plate 94 and upper plate 96. In one embodiment, the horizontal channel 98 is etched into the lower surface of the middle plate 94. In another embodiment, a portion, or all, of the flow-through channel 99 is machined out of one or more of the plates. The approximately horizontal channel 98 is connected to an approximately vertical 30 channel 97A that passes through the middle plate 94. Channel 97A is aligned with channel 97B of upper plate 96, such that the cumulative vertical channel 97A, B passes through two or more of the plurality of plates. The horizontal channel 98 is connected to inlet port INLET, and the vertical channels 97A, B is connected to outlet port OUTLET. In one embodiment, horizontal channel 98 and vertical channels 97A, B are configured to fluidly couple the inlet port INLET to the outlet port OUTLET when assembled as a neutron reflector. The upper plate 96 may comprise one or more approximately vertical channels that only passes through a single plate. In one embodiment, at least some of the plurality of plates 92, 94, 96 have different thicknesses, wherein a distance between inlet ports varies according to the different thicknesses of the plurality of plates. For example, middle plate 94 has a thickness of Dn, whereas upper plate 96 has a thickness of Di. In one embodiment, plate thickness Di associated with an upper end of the neutron reflector 90 is larger or greater than plate thickness Dn associated with a central portion of the neutron reflector 90. The number of plates are shown as being three for simplicity of illustration, whereas other embodiments could include a lesser or greater number of plates. In one embodiment, the number of plates in the neutron reflector 90 is approximately equal to the number of horizontal channels. In another embodiment, the number of plates is greater than the number of horizontal channels. In yet another embodiment, for example where multiple horizontal channels are formed between any two adjacent plates, the number of plates may be fewer than the number of horizontal channels. The construction illustrated by FIG. 9 minimizes or reduces machining costs and material waste that are associated with conventional designs. Whereas the horizontal channels are illustrated as slots or channels, the ports may comprise circular holes in certain embodiments. For example, circular inlet or outlet ports may be drilled, machined, or otherwise formed in or through the neutron reflector. FIG. 10 illustrates a partial cross-sectional top view of a power module assembly 150 including a reactor core 6 surrounded by a neutron reflector 140 about its perimeter. The neutron reflector 140 comprises a plurality of outlet ports 160 configured to output a coolant flow that has partially passed through the reactor core 6. Neutron reflector 140 may comprise a plurality of subsections 140A, 140B that are combined during assembly of the power module assembly 150. In one embodiment, each of the plurality of subsections 140A, 140B comprise a number of layered plates similar to those illustrated by the neutron reflector 90 of FIG. 9. Whereas the outlet ports 160 are illustrated as slots or channels, the ports may comprise circular holes in certain embodiments. For example, circular outlet ports may be drilled, machined, or otherwise formed in or through the neutron reflector. FIG. 11 illustrates a novel method 200 of circulating coolant through a reactor housing using a neutron reflector configured to receive coolant flow from a reactor core. The method 200 may be understood to operate with, but not limited by, various embodiments illustrated herein as FIGS. 1-10. At operation 210, a primary coolant is circulated into a reactor housing. At operation 220, the primary coolant is partitioned into a first portion and a second portion, wherein the first portion passes entirely through a reactor core located in the reactor housing, and wherein the second portion passes partially through the reactor core before entering an inlet of a neutron reflector. In one embodiment, the inlet is located in a side wall of the neutron reflector, wherein the side wall faces the reactor core. At operation 230, the second portion of the primary coolant is directed through the neutron reflector. In one embodiment, the second portion of the primary coolant enters the neutron reflector primarily as single-phase liquid, or two-phase liquid and vapor. At operation 240, the second portion of the primary coolant that exits the neutron reflector is recombined with the first portion of the primary coolant that passes through the reactor core. The second portion of the primary coolant may be recombined with the first portion within the reactor housing to increase a flow rate of the primary coolant through the reactor core. In one embodiment, the second portion of the primary coolant exits the neutron reflector as two-phase steam and liquid. In yet another embodiment, the second portion of the primary coolant exits the neutron reflector as single-phase steam. Although the embodiments provided herein have primarily described a pressurized water reactor, it should be apparent to one skilled in the art that the embodiments may be applied to other types of nuclear power systems as described or with some obvious modification. For example, the embodiments or variations thereof may also be made operable with a boiling water reactor. Dimensions of the figures are not provided to scale, and in some cases certain features have been exaggerated in scale in order to illustrate or describe certain details. The flow rates of coolant in the reactor vessel, as well as other rates and values described herein are provided by way of example only. Other rates and values may be determined through experimentation such as by construction of full scale or scaled models of a nuclear reactor. Having described and illustrated the principles in example embodiments, it should be apparent that such embodiments may be modified in arrangement and detail without departing from such principles. We claim all modifications and variation coming within the spirit and scope of the following claims. |
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claims | 1. A process for manufacturing a nano aluminum composite, comprising the steps of:a) providing an aluminum powder having a natural oxide formation layer and an aluminum oxide content between about 0.1 and about 4.5 wt. % and a specific surface area of from about 0.3 and about 5.0 m2/g, the aluminum powder having a d90 particle size of about 2.3 microns and a d10 particle size of about 0.6 microns;b) hot working the aluminum powder at a temperature below the recrystallization temperature of the powder, and forming thereby a superfine grained matrix aluminum alloy; andc) simultaneously with the hot working of step b), redistributing the aluminum powder into uniformly dispersed nano particles of alumina throughout said alloy;d) subsequent to steps b) and c), blending the superfine grained matrix aluminum alloy with a ceramic particulate to form a powder mixture, whereby the powder mixture comprises about 5 wt. % to about 40 wt. % of the ceramic particulate;wherein said superfine grained matrix aluminum alloy has an average particle size of about 200 nm. 2. The process according to claim 1, wherein step b) of hot working is carried out at a temperature less than the melting point of said alloy. 3. The process according to claim 1, wherein the aluminum powder of step a) has a d50 particle size of about 1.3 microns. 4. The process according to claim 1, wherein the ceramic particulate is selected from the group consisting of silica, silicon carbide, boron carbide, boron nitride, titanium oxide, titanium diboride, and mixtures thereof. 5. The process according to claim 4, wherein the powder mixture is sintered to form a billet. 6. The processing according to claim 1, wherein the natural layer of aluminum oxide on the aluminum powder of step a) has a thickness of between 3-7 nm. 7. The process according to claim 1, wherein said process is free of mechanical alloying. 8. A process for manufacturing a nano aluminum composite, comprising the steps of:a) providing an aluminum powder having a natural oxide formation layer and an aluminum oxide content between about 0.1 and about 4.5 wt. % and a specific surface area of from about 0.3 and about 5.0 m2/g;b) hot working the aluminum powder at a temperature below the recrystallization temperature of the powder, and forming thereby a superfine grained matrix aluminum alloy; andc) simultaneously with the hot working of step b), redistributing the aluminum powder into uniformly dispersed nano particles of alumina throughout said alloy;d) subsequent to steps b) and c), blending the superfine grained matrix aluminum alloy with a ceramic particulate to form a powder mixture, the ceramic particulate comprising boron carbide having a particle size distribution of 100% less than about 250 microns and the boron carbide is nuclear grade;e) sintering the powder mixture to form a billet;wherein said superfine grained matrix aluminum alloy has an average particle size of about 200 nm. 9. The process according to claim 8, wherein step b) of hot working is carried out at a temperature less than the melting point of said alloy. 10. The process according to claim 8, subsequent to steps b) and c), whereby the powder mixture comprises about 5 wt. % to about 40 wt. % of the ceramic particulate. 11. The processing according to claim 8, wherein the natural layer of aluminum oxide on the aluminum powder of step a) has a thickness of between 3-7 nm. 12. The process according to claim 8, wherein said process is free of mechanical alloying. 13. A process for manufacturing a nano aluminum composite, comprising the steps of:a) providing an aluminum powder having a natural oxide formation layer and an aluminum oxide content between about 0.1 and about 4.5 wt. % and a specific surface area of from about 0.3 and about 5.0 m2/g, the aluminum powder having a particle size of less than about 30 μm in diameter and the natural layer of aluminum oxide has a thickness of between 3-7 nm;b) hot working the aluminum powder at a temperature below the recrystallization temperature of the powder, and forming thereby a superfine grained matrix aluminum alloy; andc) simultaneously with the hot working of step b), redistributing the aluminum powder into uniformly dispersed nano particles of alumina throughout said alloy;d) subsequent to steps b) and c), blending the superfine grained matrix aluminum alloy with a ceramic particulate to form a powder mixture, whereby the powder mixture comprises about 5 wt. % to about 40 wt. % of the ceramic particulate;wherein said superfine grained matrix aluminum alloy has an average particle size of about 200 nm. 14. The process according to claim 13, wherein subsequent to step d), sintering the powder mixture to form a billet. 15. The process according to claim 13, wherein the ceramic particulate is selected from the group consisting of silica, silicon carbide, boron carbide, boron nitride, titanium oxide, titanium diboride, and mixtures thereof. 16. The aluminum alloy of claim 15, wherein the ceramic particulate is boron carbide having a particle size distribution of 100% less than about 250 microns and the boron carbide is nuclear grade. 17. The process according to claim 13, wherein step b) of hot working is carried out at a temperature less than the melting point of said alloy. 18. The process according to claim 13, wherein said process is free of mechanical alloying. |
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description | This patent application claims the benefit of U.S. Provisional Patent Application No. 61/484,319 filed on May 10, 2011, which is incorporated herein by reference in its entirety. 1. Field of the Invention The present invention relates to the inspection of welds using vibrothermography, and, more particularly, the present invention relates to inspecting and determining the quality of J-groove welds of a nuclear reactor pressure vessel (RPV) head and the RPV bottom mounted nozzles (BMNs) using vibrothermography. 2. Description of the Related Art Nuclear reactors have a reactor cover or closure head through which a series of openings having tubes extending therethrough to the inside surface of the reactor are welded to the closure head by J-groove welds. These tubes have control rod drives and instrumentation packages sealably extending into the reactor internals. Similarly, the bottom portion of the RPV contains penetrations with instrumentation packages that are also connected to the RPV via J-groove welds. Certain parts of the reactor vessel head are known to be susceptible to stress corrosion cracking. Many reactor heads have been completely replaced well before the planned design life because of extensive cracking in the nozzles and/or J-groove welds. The J-groove welds must be inspected during reactor shut down for any cracks that may have developed in the welds. Such inspections are usually made from beneath the reactor head with the head on a supporting head-stand during normal refueling periods. J-groove welds are difficult to inspect due to the complex geometry and high radiation fields under the head. Visual techniques are known to have difficulty with very small, tight surface cracks, especially on rough surfaces or after certain processes (i.e., grinding are used to improve surface finish for inspection. Standard nondestructive examination techniques such as eddy current and dye-penetrant examinations are the typical inspection approach used; however, application of these techniques is time-consuming in an operational environment where delays associated with inspections are undesirable. If ultrasound techniques are used for an RPV head inspection, the water couplant must be managed and disposed of following the inspection. Dye-penetrant techniques also suffer from disadvantages including requirements for cleaning the surface, a process that can involve significant volumes of solvent and chemicals. High radiation under the reactor head makes it undesirable for personnel to perform these inspections by direct contact with the components; thus, various robotic elements are used to remotely access these welds with the mentioned nondestructive testing machinery. The robotic challenges with these techniques are significant, typically requiring a close proximity or actual contact to the tested weld surface which is difficult to achieve with the robotic delivery devices and the remotely controlled tools used to access the welds. Thus, some form of nondestructive remote inspection of J-groove welds that did not require close proximity or contact with the welded surface, nor the exact robotic manipulation associated therewith (which is difficult to implement), was needed. One such form of testing is described in U.S. Pat. No. 6,856,662, the disclosure of which is incorporated herein by reference. This patent describes a photothermal inspection method, which includes sequentially heating small areas of a weld using a laser while monitoring the weld area using a remote infrared camera. The laser and camera are positioned at an angle of approximately 45° to the inspection surface normal. At cracks or other anomalies (referred to herein collectively as cracks), the thermal wave imparted by the laser does not propagate as rapidly, thereby producing an indication of the crack as a sharp drop in the thermal scan at the point of the crack. The temperature differential is detected using the infrared camera. While the photothermal inspection method of the '662 patent is an improvement over prior J-groove weld inspection methods, it is desirable to use a nondestructive inspection system and method that allows for a quicker inspection. The present invention is related to a system and method of inspecting the J-groove welds of a nuclear reactor vessel head or BMN using vibrothermography. The weld to be inspected is subjected to a transient sonic excitation while the weld area is monitored using a remote infrared camera. The sonic excitation induces mechanical vibration, which causes heat generation at any cracks in the weld. The infrared camera detects any temperature differentials in the weld, indicating the presence of a crack. For underwater examinations, as in the case of a BMN, a cofferdam may be used to purge the water surrounding the J-groove weld to make the inspection process possible. This vibrothermographic inspection is quicker than photothermal inspection. The vibrothermography system and method utilizes much larger inspection areas, eliminating the need to work incrementally with small inspection areas. Vibrothermographic inspection is a form of active thermography wherein the applied external stimulus causes the defect itself to generate heat. The sonic excitation stimulus of vibrothermographic inspection is effective regardless of the angle of attack to the weld. Likewise, detection of the crack's heat signature is largely independent of the camera angle relative the inspection zone. Thus, the setup for vibrothermographic inspection is easier and quicker. For all of these reasons, vibrothermographic inspection is an improvement over known inspection techniques. Vibrothermography is also not sensitive to the surface roughness of an inspected object (as long as there is a line of sight to the inspected areas) and vibrothermography can deal with complex geometries/surfaces much better than ultrasound or eddy current since the camera is an optical device only requiring line-of-sight access to the target surface. The method of inspecting a J-groove weld in a nuclear reactor may include providing a vibration device and an imaging device near the weld. The vibration device may make physical contact with the weld to be inspected, but more preferably does not make contact with the weld. The vibration device is used to impart acoustic energy to the weld under inspection, and may impart one or more of a range of frequencies to the weld. The range may extend from near DC frequencies to over 50 kHz, and preferably extends to over 100 kHz. The imparted acoustic energy is chosen so as to generate vibration of the weld and friction at a flaw that may be located within the weld, which friction generates heat. The flaw may be located beneath the surface of the weld, in which case the generated heat may be transferred by conduction to the surface of the weld. The method further includes obtaining a thermal image of the weld using the imaging device such as an infrared camera. The image(s) captured by the imaging device is processed and analyzed to determine the presence of any heat differential within the image, such as an area of relatively higher temperature caused by vibrational friction at a flaw. The determination can then be made, based on the analysis, whether a flaw within the weld or heat affected zone is present. The inspection method may be repeated a number of times to inspect multiple J-groove welds within the reactor. As the vibration device need not be in contact with the weld under inspection, and furthermore need not be positioned individually with respect to the weld under inspection, the inspections of multiple J-groove welds could be made without repositioning the vibration device. The vibration device preferably includes a broadband or single frequency piezoelectric transducer to induce vibration of the weld, the piezoelectric transducer preferably being variable to impart a range of frequencies and vibration amplitudes to the weld. The vibration energy transmitted by the vibration device is chosen so as not to exacerbate any flaw within the weld under inspection or any other welds in the proximity of the inspection system. The inspection system may further include a manipulator to which the imaging device is coupled, and a communication system operatively linking the vibration device, the imaging device, and the manipulator. The system can thus be remotely controlled to position the camera and/or manipulator as needed to perform inspections. The vibration device may also be coupled to the manipulator, allowing the user to reposition it as needed. The communication system preferably is software-driven, synchronizing the inspection system components. Vibrothermography, also known as sonic IR and thermosonics, is a nondestructive evaluation technique that uses sonic excitation (vibration) to preferentially heat cracks in metal, ceramics, polymers, and composites. This local temperature change may be detected remotely using an infrared (IR) camera. This phenomenon and general test approach is applied to the reactor vessel head by injecting a transient sonic excitation while monitoring the J-groove weld area with a remote infrared camera. Any cracks that “light-up” can be detected and marked for further characterization and evaluation or repair. The instant system and inspection method provides advantages over other known vibrothermography or alternative non-destructive examination inspection systems and methods. The inventive technique can be implemented with simple remote tools that do not require complex motions to track the complex weld surface. The crack indication is visually clear and can readily be correlated with the exact position on the structure (unlike, for example, eddy current examination techniques). The camera can be operated and positioned remotely from the weld under inspection, reducing radiation influences on the camera. Unlike penetrant or ultrasonic techniques, there are no chemicals or solvents to manage or dispose of following the procedure. Vibrothermography requires little or no surface cleaning, and can be performed significantly faster than known techniques such as ultrasonic, penetrant, and eddy current examinations. The rate of false calls (a defect is called when none is present) for vibrothermography is lower than with most other nondestructive examination techniques. See, for example, J. DiMambro et al., “Sonic Infrared (IR) Imaging and fluorescent Penetrant Inspection Probability of Detection (POD) Comparison,” American institute of Physics, AIP Conf. Proc. 894, pp. 463-470, 2006. A known nuclear reactor vessel 1 is shown in FIG. 1, and an exemplary head 10, tube 12, and J-groove weld 14 configuration is shown in FIG. 2. The vessel 1 has a series of Inconel 600 control rod drive tubes 12 extending through the dome of the reactor head 10. The carbon steel dome is clad with approximately 0.5 in. of stainless steel. The tubes 12 are welded to the head 10 by a known J-groove weld 14, which must be inspected for flaws during reactor refuel and maintenance outages to ensure the weld integrity remains intact. FIG. 3 shows a preferred layout for vibrothermographic inspection of the J-groove welds 14. The head 10 has been removed from the reactor vessel 1 and placed on a supporting head-stand 16, such as during a normal refueling outage. One or more sonic exciters 20 have been positioned at advantageous positions near the head 10 and/or tubes 12. The exciters 20 can be in contact with the head 10, or tubes 12, and need not be in direct contact with the welds 14. While three such exciters 20 are show in the illustrative embodiment depicted in FIG. 3, more or fewer exciters 20 can be used. Likewise, the exciter locations shown in FIG. 3 are exemplary; other locations could be used as determined by the user for the particular application. Application of acoustic energy in the instant invention is novel with respect to known vibrothermographic systems. One and often more broadband piezoelectric transducers are preferably used with the instant invention and have been modified to function underwater. The inventive excitation system has significantly improved functionality over the ultrasonic welders, piezoelectric transducers, and variations thereof used in nearly all other vibrothermographic inspections. An important aspect of nondestructive testing is that the testing not create or worsen any existing damage in the equipment being inspected. Typically-used ultrasonic welders have been shown to grow cracks, and thus fail to be nondestructive. Moreover, this crack growth can occur quite rapidly, on the order of 10% to 500% or more growth after approximately 10-15 seconds of excitation. Thus, typical vibrothermographic inspection systems are not appropriate for use in environments where equipment integrity and safety are important, such as a nuclear power plant. The instant excitation system is a lower-power system that does not negatively affect existent cracks. The instant weld inspection system and method is therefore appropriate for use in environments where safety is a primary concern. The instant excitation system is also variable, providing a wider bandwidth of imparted vibration. Whereas most known systems have a bandwidth of 15-25 kHz or a single frequency at, typically, 20, 30, or 40 kHz, the exciters 20 preferably are broadband transducers capable of generating vibrations over a 50 kHz range of frequencies, and more preferably having a range of 100 kHz or more, from near DC to more than 100 kHz. The exciters 20 are robust, and have been modified to allow them to function in moist, humid, condensing, or underwater environments. This modification may be achieved, for example, by coating the exciters 20 is an adhesively-bonded polymer coating. Other means of such modification will be apparent to those of skill in the art. Such modification is particularly useful in the inspection of J-groove welds 14 of a nuclear reactor vessel head 10 and allows some inspections to be performed in a noncontact setup by coupling vibrations to the inspected structure through water. This functionality is not present in any other known vibrothermography systems. A cofferdam 30 may be used to purge the water surrounding the J-groove to facilitate the inspection process. FIG. 6 shows a possible layout of for vibrothermographic inspection system of the present invention of a BMN 18. An air hose 32 and valve 34 may be used in conjunction with the cofferdam 30 to replace water within the cofferdam 30 with air. It may not be necessary to remove all of the water from within the cofferdam 30, as shown in the illustrated embodiment of FIG. 6. Additional benefits of the instant exciters 20 include higher system repeatability and more compact size allowing fur inspections in confined spaces. In use, the exciters 20 are programmed to impart vibration energy to the tubes 12 and J-groove welds 14. The frequency of the imparted energy is a function of the target components, and includes such considerations as material properties and dimensions of such components. Preferably, a range of frequencies is imparted from the exciters 20 to the tubes 12 and welds 14. The imparted energy causes the target components to vibrate. This vibration causes friction at any crack within the welds 14, which friction generates heat. This localized heat signature is detected by the infrared camera 22, which is positioned to observe one or more of the tubes 12 and welds 14. The instantaneous detection zone of the instant system and method is much larger than with a photothermal inspection method. For example, the detection zone of the instant invention may fully inspect a square foot or larger area in a matter of a few (i.e., less than five) seconds, whereas the detection zone of a photothermal system during the same amount of time may be a ten square inches or less. As the presence of heat indicates the presence of a crack, the instant system and method can detect cracks that are internal to the weld 14 by the heat transfer from the crack outward to the surface of the weld 14. FIGS. 4 and 5 illustrate an example measurement. FIG. 4 shows a raw thermal image of a flawed weld sample. While not seen in this figure, the location of a bad weld is highlighted. by the box. An image of the same weld sample after being subjected to vibration from an exciter 20 is shown in FIG. 5. This crack is readily detected by its generated thermal signature and associated image processing. The camera is coupled to a robotic manipulator 24 that is used to position the camera 22 as needed to make inspections of the various J-groove welds 14. As the exciters 20 do not have to be repositioned for the inspection of each weld 14, the exciters 20 do not have to be coupled to the manipulator 24. The user may choose to couple one or more exciters 20 to the manipulator 24, however, which may eliminate the need for a separate step of positioning the exciters 20 within the head 10. In some instances, it may be possible to detect the thermal signatures of many cracks from a significant distance, such as from outside the refueling pool in which the head 10 is positioned during a refueling outage. In such an instance, it may be possible for a person to manually handle the camera safely, foregoing the need to use the robotic manipulator 24, at least for inspection of the upper portion of the welds 14 (near the outer surface of the head 10). Each of the exciter 20, camera 22, and manipulator 24 (if used) preferably is remotely controlled. Preferably, each of these components is controlled by software that is either fully automated or user-driven. The control system allows for individual or multiple exciters 20 to be used, and for single frequency, multiple simultaneous frequencies, and/or frequency sweep excitation modes to be used. If multiple exciters 20 are used, they may be programmed to use the same or different frequencies or frequency ranges. The control system can also process the image data captured by the camera 22 to pick out important information from each test. This software can be expanded as necessary to incorporate changing technical requirements. While the preferred embodiments of the present invention have been described above, it should be understood that they have been presented by way of example only, and not of limitation. It will be apparent to persons skilled in the relevant art that various changes in form and detail can be made therein without departing from the spirit and scope of the invention. Thus the present invention should not be limited by the above-described exemplary embodiments, but should be defined only in accordance with the following claims and their equivalents. Furthermore, while certain advantages of the invention have been described herein, it is to be understood that not necessarily all such advantages may be achieved in accordance with any particular embodiment of the invention. Thus, for example, those skilled in the art will recognize that the invention may be embodied or carried out in a manner that achieves or optimizes one advantage or group of advantages as taught herein without necessarily achieving other advantages as may be taught or suggested herein. |
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claims | 1. A respiration phantom for performing quality assurance on a radiation delivery system, comprising:a human-like skeletal structure;at least one deformable component positionable at least partially internal to the human-like skeletal structure, the deformable component having a shape resembling an organ of a human anatomy, wherein the deformable component attenuates radiation substantially similarly to the organ of the human anatomy; anda respiration actuator positioned to deform the deformable component with a respiration-like motion. 2. The respiration phantom of claim 1, wherein the human-like skeletal structure is radiographically distinct, wherein the respiration phantom comprises an anthropomorphic phantom that radiographically images substantially similar to the human anatomy, and wherein the respiration phantom attenuates radiation substantially similar to the human anatomy. 3. The respiration phantom of claim 2, wherein the deformable component is removable from the skeletal structure. 4. The respiration phantom of claim 3, wherein the deformable component comprises a radiologically sensitive gel in a deformable container shaped to resemble the organ of the human anatomy. 5. The respiration phantom of claim 4, wherein the organ of the human anatomy comprises a lung and wherein the deformable container comprises a lung shaped container. 6. The respiration phantom of claim 3, wherein the human-like skeletal structure includes a rib cage and further comprising a plurality of deformable components positionable internal to the human-like skeletal structure, the plurality of deformable components each having a different shape resembling a different organ of the human anatomy, wherein the plurality of deformable components are removeable from the human-like skeletal structure and wherein the plurality of deformable components each attenuate radiation substantially similar to a corresponding organ of the human anatomy. 7. The respiration phantom of claim 6, further comprising:a thoracic cavity formed within the human-like skeletal structure; anda diaphragm positioned within the human-like skeletal structure, wherein the respiration actuator is coupled to move the diaphragm along an inferior to superior axis and wherein the thoracic cavity is flexible to expand along a posterior to anterior axis in response to motion from the diaphragm. 8. The respiration phantom of claim 6, further comprising a skin-like sheath surrounding the human-like skeletal structure. 9. The respiration phantom of claim 6, wherein the plurality of deformable components comprise foam organs, and wherein each of the plurality of deformable components are each individually removable and replaceable with a radiologically sensitive gel organ. 10. The respiration phantom of claim 6, wherein the respiration actuator comprises a mechanical actuator coupled to reciprocally compress the plurality of deformable components along an inferior to superior axis and to cause the rib cage to expand along a posterior to anterior axis. 11. The respiration phantom of claim 6, wherein the respiration actuator comprises a pneumatic actuator coupled to reciprocally compress the plurality of deformable components along an inferior to superior axis and to cause the rib cage to expand along a posterior to anterior axis. 12. The respiration phantom of claim 1, wherein the respiration actuator is programmable to simulate multiple different respiration-like motions. 13. A system, comprising:a respiration phantom comprising:a human-like skeletal structure;at least one deformable component positionable at least partially internal to the human-like skeletal structure, the deformable component having a shape resembling an organ of a human anatomy, wherein the deformable component attenuates radiation substantially similarly to the organ of the human anatomy; anda respiration actuator positioned to deform the deformable component with a respiration-like motion; anda radiation delivery system including a radiation source and a patient positioning system, the patient positioning system to position the respiration phantom at a preset position and the radiation source moveable to deliver a dose of radiation to the respiration phantom while the respiration actuator is deforming the deformable component with the respiration-like motion. 14. The system of claim 13, wherein the radiation delivery system comprises an image guided radiation delivery system, wherein the image guided radiation delivery system further includes an imaging system to image the respiration phantom and provide real-time feedback to align the radiation source during delivery of the dose of radiation. 15. The respiration phantom of claim 14, wherein the human-like skeletal structure is radiographically distinct, wherein the respiration phantom comprises an anthropomorphic phantom that radiographically images substantially similar to the human anatomy, and wherein the respiration phantom attenuates radiation substantially similar to the human anatomy. 16. The system of claim 15 wherein the respiration phantom includes a sensor array within the human-like skeletal structure to determine whether the dose of radiation was delivered to the respiration phantom as expected. 17. The system of claim 15, wherein the deformable component is removable from the skeletal structure for analysis to determine whether the dose of radiation was delivered to the deformable component as expected. 18. The system of claim 17, wherein the deformable component comprises a radiologically sensitive gel in a deformable container shaped to resemble the organ of the human anatomy. 19. The system of claim 17, wherein the human-like skeletal structure includes a rib cage and further comprises a plurality of components positionable internal to the human-like skeletal structure, the plurality of components each having a different shape resembling a different organ of the human anatomy, wherein the plurality of components are removeable from the human-like skeletal structure and wherein the plurality of components each attenuate radiation substantially similar to a corresponding organ of the human anatomy. 20. The system of claim 19, wherein the respiration actuator is coupled to reciprocally compress the plurality of components along an inferior to superior axis and to cause the rib cage to expand along a posterior to anterior axis. 21. A respiration phantom for performing quality assurance on a radiation delivery system, comprising:means for providing a human-like skeletal structure;means for attenuating radiation substantially similar to a organ of a human anatomy, the means for attenuating positioned internal to the means for providing a human-like skeletal structure, the means for attenuating having a shape resembling the organ of a human anatomy and being deformable; andactuator means for deforming the means for attenuating with a respiration-like motion. 22. The respiration phantom of claim 21, wherein the means for providing a human-like skeletal structure is radiographically distinct, wherein the respiration phantom comprises an anthropomorphic phantom that radiographically images substantially similar to the human anatomy, and wherein the respiration phantom attenuates radiation substantially similar to the human anatomy. 23. The respiration phantom of claim 22, wherein the means for attenuating is removable from the means for providing a human-like skeletal structure. 24. The respiration phantom of claim 23, wherein the means for attenuating comprises a radiologically sensitive gel in a deformable container shaped to resemble the organ of the human anatomy. 25. The respiration phantom of claim 22, further comprising a plurality of means for attenuating positionable internal to the means for providing the human-like skeletal structure, the plurality of means for attenuating each having a different shape resembling a different organ of the human anatomy, wherein the plurality of means for attenuating are removeable from the means for providing the human-like skeletal structure and wherein the plurality of means for attenuating each attenuate radiation substantially similar to a corresponding organ of the human anatomy. 26. The respiration phantom of claim 25, wherein the actuator means comprises a means to reciprocally compress the plurality of means for attenuating along an inferior to superior axis of the respiration phantom and to cause the means for providing the human-like skeletal structure to expand along a posterior to anterior axis of the respiration phantom. 27. A method of performing quality assurance on a radiation delivery system, comprising:cyclically compressing a deformable component internal to a respiration phantom having a human-like skeletal structure along an inferior to superior axis of the respiration phantom, the deformable component having a shape resembling an organ of a human anatomy, wherein the cyclical compression causes the human-like skeletal structure to cyclically expand along a posterior to anterior axis of the respiration phantom; andemitting a radiation beam from a radiation source of the radiation delivery system at the respiration phantom while cyclically compressing the deformable component. 28. The method of claim 27, further comprising analyzing a dose of radiation delivered to the respiration phantom to determine whether the radiation delivery system is calibrated. 29. The method of claim 27, further comprising positioning the respiration phantom at a preset position with a robotic couch positioning system. 30. The method of claim 29, wherein positioning the respiration phantom comprises positioning the respiration phantom under feedback guidance of a radiation image guidance subsystem of the radiation delivery system. 31. The method of claim 30, wherein the human-like skeletal structure is radiographically distinct, wherein the respiration phantom comprises an anthropomorphic phantom that radiographically images substantially similar to the human anatomy, and wherein the respiration phantom attenuates radiation substantially similar to the human anatomy. 32. The method of claim 27, wherein emitting the radiation beam at the respiration phantom comprises emitting the radiation beam at the deformable component, and further comprising:removing the deformable component from the respiration phantom, wherein analyzing the dose of radiation delivered to the respiration phantom comprises analyzing the dose of radiation delivered to the deformable component to determine whether the radiation delivery system is calibrated. 33. The method of claim 32, wherein the deformable component comprises a radiologically sensitive gel in a deformable container shaped to resemble the organ of the human anatomy. |
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claims | 1. A transport container for a nuclear fuel assembly of elongate shape in a longitudinal direction, the container comprising:a support having at least a first longitudinal bearing surface delimiting a longitudinal housing for receiving a nuclear fuel assembly;a door having a second longitudinal bearing surface, the second longitudinal bearing surface including a pair of adjacent faces meeting at a vertex to form an angle, the door being movable between a holding position holding the nuclear fuel assembly between the first and second longitudinal bearing surfaces and a release position in which the nuclear fuel assembly is not held between the first and second longitudinal bearing surfaces; andan adjuster adjusting a transverse spacing between the first and second longitudinal bearing surfaces when the door is in the holding position. 2. The container as recited in claim 1, wherein the first longitudinal bearing surface comprises a first pair of adjacent longitudinal faces, and the second longitudinal bearing surface comprises a second pair of adjacent longitudinal faces, each of the second pair of adjacent longitudinal faces parallel with and opposite a corresponding one of the longitudinal faces of the first pair of adjacent longitudinal faces when the door is in the holding position. 3. The container as recited in claim 2 wherein the first and second pairs of adjacent longitudinal faces converge towards first and second vertices, respectively, and the adjuster comprises a means for adjusting a position of the door with respect to the support by translation of the door in a transverse adjusting direction when the door is in the holding position. 4. The container as recited in claim 3 wherein the support comprises parallel longitudinal surfaces for guiding the translation of the door in the adjusting direction. 5. The container as recited in claim 3 wherein the container comprises means for displacing the door with respect to the support between the holding position and the release position by translation in the direction of adjustment and then rotation about at least one longitudinal shaft coupled to the support and to the door. 6. The container as recited in claim 2 wherein the first pair of adjacent longitudinal faces form an angle equal to an angle formed by the second pair of adjacent longitudinal faces, the angle being from 60° to 135°. 7. The container as recited in claim 1 wherein the second longitudinal bearing surface is free from movable runners for resting on a nuclear fuel assembly. 8. The container as recited in claim 1 for transporting a nuclear fuel assembly. 9. The container as recited in claim 8 wherein the container is used with the same support and the same door to transport nuclear fuel assemblies of at least two different types. 10. The container as recited in claim 2 wherein the second pair of adjacent longitudinal faces connect with each other at a vertex to form an angle. 11. The container as recited in claim 1 wherein the angle is a right angle. 12. The container as recited in claim 1 wherein the first longitudinal bearing surface includes a further pair of adjacent faces meeting at a further vertex to form a further angle. 13. A transport container for a nuclear fuel assembly of elongate shape in a longitudinal direction, the container comprising:a support having at least a first longitudinal bearing surface delimiting a longitudinal housing for receiving a nuclear fuel assembly;a door extending over an entire longitudinal length of the housing, the door having a second longitudinal bearing surface, the door being movable between a holding position holding the nuclear fuel assembly between the first and second longitudinal bearing surfaces and a release position in which the nuclear fuel assembly is not held between the first and second longitudinal bearing surfaces; andan adjuster adjusting a transverse spacing between the first and second longitudinal bearing surfaces when the door is in the holding positionwherein the second longitudinal bearing surface is free from movable runners for resting on a nuclear fuel assembly. 14. The container as recited in claim 1 wherein the adjuster includes screws rotatably mounted to nuts. 15. The container as recited in claim 14 wherein the nuts are connected to the door and the screws are rotatable with respect to the support via bearings connected to the support. 16. The container as recited in claim 1 wherein the adjuster moves the door translationally so that the second longitudinal surface moves translationally with respect to the first longitudinal surface. 17. The container as recited in claim 13 wherein the second longitudinal bearing surface is capable of resting directly on the nuclear fuel assembly. 18. A transport container for a nuclear fuel assembly of elongate shape in a longitudinal direction, the container comprising:a support having at least a first longitudinal bearing surface delimiting a longitudinal housing for receiving a nuclear fuel assembly;a door extending over an entire longitudinal length of the housing, the door having a second longitudinal bearing surface, the door being movable between a holding position holding the nuclear fuel assembly between the first and second longitudinal bearing surfaces and a release position in which the nuclear fuel assembly is not held between the first and second longitudinal bearing surfaces; andan adjuster adjusting a transverse spacing between the first and second longitudinal bearing surfaces when the door is in the holding position;wherein the first longitudinal bearing surface comprises a first pair of adjacent longitudinal faces, and the second longitudinal bearing surface comprises a second pair of adjacent longitudinal faces, each of the second pair of adjacent longitudinal faces parallel with and opposite a corresponding one of the longitudinal faces of the first pair of adjacent longitudinal faces as the adjuster adjusts a transverse spacing between the first and second longitudinal bearing surfaces when the door is in the holding position;wherein the second pair of adjacent longitudinal faces connect with each other at a vertex to form an angle. 19. The transport container recited in claim 1 wherein the support includes a bottom surface and the first longitudinal bearing surface includes a pair of support faces angled with respect to each other in a V-shape, the support faces extending away from the bottom surface. 20. The transport container recited in claim 19 wherein the first longitudinal bearing surface and the second longitudinal bearing surface define a volume in the holding position, the adjuster being configured to move the door toward and away from the bottom surface in the holding position such that the volume increases and decreases and the vertex of the second longitudinal bearing surface moves toward the first longitudinal bearing surface. |
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description | This application is a Divisional Application of U.S. application Ser. No. 10/228,139 filed Aug. 27, 2002, now U.S. Pat. No. 6,953,930, and based upon and claims the benefit of priority to Japanese Patent Application Nos. 2001-256206, filed Aug. 27, 2001; and 2001-341651, filed Nov. 7, 2001, the entire contents each of which are incorporated herein. 1. Field of the Invention The present invention relates to a conductive transparent probe and a probe control apparatus. More particularly, the present invention relates to a conductive transparent probe used in a tunneling luminescence microscope, and a probe control apparatus for controlling a distance between the apex of a probe and a sample, wherein the tunneling luminescence microscope measures optical and electronic characteristics of a very small region of a size of the nanometer order by detecting luminescence caused by applying a probe current into the sample. 2. Description of the Related Art As devices become small and technologies for utilizing characteristics of individual molecules develop, great demands have arisen for technologies for characteristic evaluation of a very small region of a size of the nanometer order in materials (to be referred to as a nano region hereinafter), and for technologies for optical and electronic characteristic measurement of individual molecules intrinsically having a size of the nanometer order. For realizing such measurement and evaluation, a tunneling luminescence microscope (to be referred to as a TL microscope hereinafter) is provided that enables detection and analysis of luminescence caused by applying a current from an apex of a sharpened probe to a sample. In addition, a probe that is transparent and has conductivity (to be referred to as a conductive transparent probe hereinafter) has been developed, wherein the conductive transparent probe applies a current from its apex into a sample, and at the same time, receives and collects luminescence from the apex, so that luminescence collection yield is improved. The conductive transparent probe is powerfully used for characteristic evaluation of a nano region. As effectiveness of the TL apparatus for characteristic evaluation of a nano region increases, it is demanded by users that the sample to be measured is not only a material having only a conductive region but also a material in which a nonconductive region or a highly resistive region is mixed with the conductive region. In an apparatus (to be referred to as a probe microscope hereinafter) that measures a sample by bringing a probe extremely close to the surface of the sample, it is very important to properly control a very small distance (to be referred to as a gap hereinafter) between the apex of the probe and the surface of the sample. Therefore, generally, as for the probe microscope (for example, a scanning tunneling microscope (to be referred to as an STM, hereinafter)) that utilizes a tunneling current flowing between the probe and the sample for measurement, a method of detecting the tunneling current flowing between the probe and the sample is used for controlling the gap (this control method is called an STM control method hereinafter). The reason for using this method for realizing precision gap control is that the tunneling current is very sensitive to the gap. However, the STM control method can be applied only to a sample of which the whole region is electronically conductive, and the STM control method cannot be applied to a sample in which a nonconductive region or a highly resistive region is mixed. Therefore, a TL apparatus that enables gap control without using the tunneling current is desperately desired, such that the TL apparatus can be applied to a sample in which a nonconductive region or a highly resistive region is mixed. As a gap control method without using the tunneling current, there is a method for utilizing an atomic force such as attractive force and repulsive force between the apex of the probe and the sample. In this method, when the apex of the probe approaches very close to the surface of the sample, atomic force between the apex and the surface is detected, and the gap is adjusted such that the detected value becomes constant. For feeding back the detected value for performing gap control, there is a method of using an optical lever and a soft probe of a cantilever shape. In this case, a laser beam is used for detecting a very small displacement of the probe. However, since the laser beam is extremely stronger than a detected signal light used for observing the sample, there is a problem in that the SN ratio decreases when measuring weak luminescence caused by the tunneling current. It is desirable to use a leaner probe made of an optical fiber in order to suppress optical transmission loss in the probe. However, it is difficult to use such a probe as the soft probe of a cantilever shape that is necessary for realizing an optical lever. In addition, there is a method called a shear force gap control method. In the method, a linear probe perpendicular to the surface of the sample is vibrated in a direction perpendicular to a center axis of the probe, so that atomic force is detected by measuring amplitude of the probe vibrating at a specific frequency. In this method, when a voltage is applied between the apex of the probe and the sample for causing luminescence, a current flows into a sensor used for detecting the amplitude, so that a detected signal is disturbed and gap control becomes unstable. Therefore, there is a problem in that a voltage cannot be applied between the probe and the sample when the shear force gap control method is used. An object of the present invention is to provide a conductive transparent probe that is applicable to the shear force gap control method while the tunneling current can be applied to a very small region without decreasing luminescence collection yield. In addition, another object of the present invention is to provide a probe control apparatus for applying a voltage between the apex of the probe and a sample so as to apply a current from the apex of the probe to cause luminescence from the sample, wherein the probe control apparatus is applicable to a sample in which a nonconductive region or a highly resistive region is mixed with a conductive region, a conductive transparent probe can be used as a probe, and it is not necessary to use a laser beam that decreases the SN ratio when weak luminescence caused by tunneling current is measured. The above-mentioned object is achieved by a conductive transparent probe used in a probe control apparatus for adjusting a distance between the apex of the conductive transparent probe and a sample by vibrating the conductive transparent probe with a vibrator in a direction perpendicular to the axis of the conductive transparent probe, the conductive transparent probe includes: an optical fiber having a taper part at one end; a conductive transparent film formed on the surface of the taper part; a first metal film formed on the surface of the optical fiber other than the taper part; wherein the conductive transparent film and the first metal film are electrically connected, and length and thickness of the first metal film are determined such that the conductive transparent probe vibrates while contacting with the vibrator. According to the above-mentioned conductive transparent probe according to the present invention, shear force gap control can be performed without losing functions of applying a probe current and collecting luminescence, and measurement by using luminescence can be performed stably even for a sample in which a nonconductive region or a highly resistive region is mixed with a conductive region. The above object is also achieved by a probe control apparatus including: a probe that is straight and vertical with respect to a surface of a sample; a vibrator for vibrating the probe in a direction perpendicular to a center axis of the probe; an amplitude detection part for detecting an amplitude of the probe; a part for controlling a distance between the apex of the probe and the sample by controlling the amplitude of the probe vibrating at a specific frequency to be a predetermined amplitude; a voltage applying part for applying a voltage between the apex of the probe and the sample; wherein the probe has optical transparency and electrical conductivity, and the probe is electrically insulated from the amplitude detection part. According to the above-mentioned probe control apparatus according to the present invention, gap control between the probe and the sample can be performed stably even for a sample in which a nonconductive region or a highly resistive region is mixed with a conductive region, for which sample it is difficult to perform gap control by using probe current. In the following, embodiments of the present invention will be descried with reference to figures. In the embodiments, a probe control apparatus will be described first, and details of a conductive transparent probe applicable to the probe control apparatus will be described next. (Probe Control Apparatus) FIG. 1 shows a block diagram of the probe control apparatus of the present invention. As shown in the figure, the probe control apparatus of the present invention includes a probe 1, a vibrator 7, a power source for vibration 9, a sensor 10, a sensing signal processing circuit 11, a sample position driving mechanism 12, a control circuit 13 for the sample position driving mechanism, a probe bias power source 14, a conductive holding plate 15 and a supporting structure 16. The apex of the linear probe 1 mounted perpendicular to the surface of the sample 3 is tapered to a point. The probe 1 is made of a linear optical fiber having optical transparency. A conductive film is applied on the periphery and the taper part of the optical fiber to provide conductivity, wherein the conductive film applied on the taper part is optically transparent for providing a luminescence collection ability. The probe 1 is held by the conductive holding plate 15 apart from the apex of the probe by 1–2 cm, so that the probe 1 is mounted on the supporting structure 16. The conductive holding plate 15 is connected to one end of the probe bias power supply 14, and supplies a current from the probe bias power supply 14 to the probe 1. The other end of the probe bias power supply 14 is connected to the sample 3, so that a voltage applying mechanism for applying voltage between the apex of the probe 1 and the sample 3 is formed. The vibrator 7 for vibrating the probe 1 in a direction perpendicular to the center axis of the probe 1 is provided on the supporting structure 16. The vibrator 7 is connected to the power supply for vibration 9, and pushes a point apart from the apex of the probe 1 by several millimeters via a sensor 10 that is an amplitude detection means, so that the probe 1 is vibrated in the direction parallel to the surface of the sample 3. The sensor 10 detects the amplitude of vibration of the probe 1, and outputs a voltage value in proportion to a displacement amount (amplitude). The output from the sensor is transmitted to the sample position driving mechanism 12 via the sensing signal processing circuit 11 and the control circuit 13. The sample position driving mechanism 12 receives an output from the control circuit 13, and moves the sample. The sensing signal processing circuit 11, the control circuit 13 and the sample position deriving mechanism 12 form a distance control means for controlling a distance (gap 4) between the apex of the probe 1 and the sample 3. An insulator 81 electrically insulates the probe 1 from the sensor 10, and an insulator 82 electrically insulates the sensor 10 from the vibrator 7. An operation of the probe control apparatus of the present invention is as follows. The probe 1 is placed on the sensor 10, and is vibrated by the sensor 10 in the direction perpendicular to the center axis of the probe 1 at a resonance frequency. The sensor 10 outputs a voltage corresponding to vibration of the probe 1. At frequencies near the resonance frequency, if the frequency changes slightly, the amplitude of the probe 1 changes greatly. Thus, the amplitude of the probe 1 is monitored with the vibrator 10 for sensing at a frequency slightly apart from the resonance frequency. In this status, the apex of the probe 1 approaches the surface of the sample 3 (the operation for the probe 1 approaching the sample 3 is referred to as “approach”). Even after starting the approach, while the gap 4 is so large that atomic force between the probe 1 and the sample 3 can be neglected, the probe 1 continues to vibrate at a constant frequency and a constant amplitude. Therefore, the amplitude of voltage output from the sensor 10 is constant, since the voltage change corresponds the vibration. Next, when the probe 1 further approaches the surface of the sample 3 so closely that atomic force becomes large, the atomic force acts as a resistance (a shear force) to the vibration of the probe 1, and the frequency changes. Therefore, the amplitude monitored by the sensor 10 changes. When the amplitude of the probe 1 changes, output voltage data of the sensor 10 also change. When the amplitude of the output voltage becomes a predetermined value, approach of the probe is stopped. After that, the gap 4 between the probe 1 and the surface of the sample is adjusted by performing feedback control such that the amplitude of the probe 1 is constant. If a current flows to the sensor 10 from the probe 1 when applying a voltage between the apex of the probe 1 and the sample 3, the detected signal output from the sensor 10 is disturbed and gap control becomes unstable. For preventing this signal disturbance, the insulator 81 is inserted between the probe 1 and the sensor 10, so that they are electrically insulated. The atomic force occurs irrespective of whether the sample 3 is conductive or nonconductive. Therefore, gap control between the probe 1 and the surface of the sample 3 can be performed even when the sample 3 includes both a conductive region and a nonconductive region or a highly resistive region. For example, when the probe 1 is placed above the nonconductive region of the sample 3, the gap 4 between the probe 1 and the surface of the sample 3 is controlled properly by using atomic force although the probe current does not flow. When the probe 1 is placed above the conductive region of the sample 3, gap control is performed by using the atomic force, and, in addition, tunneling current and luminescence caused by the current can be detected since the probe current can be applied. That is, tunneling current and luminescence caused by the tunneling current can be measured even for a sample in which a nonconductive region or a highly resistive region is mixed with a conductive region, for which sample it is difficult to control the gap 4 by using the tunneling current. In addition, since detection of the gap 4 is performed by the sensor 10 so that a laser beam is not used, the SN ratio of the detected signal light is not lowered when measuring weak luminescence caused by the tunneling current. Therefore, measurement with a high SN ratio can be achieved. In addition, since it is not necessary to use a soft probe of a cantilever shape, a probe made of a linear optical fiber applicable to forming a conductive transparent probe can be used. Further, since the insulator 81 is inserted between the probe 1 and the sensor 10 so as to electrically insulate the sensor 10 from the probe 1, current does not flow to the sensor 10 from the probe 1 even when a voltage is applied between the probe 1 and the sample 3. Thus, the detected signal is not disturbed, so that gap control is performed stably. FIGS. 2A and 2B show a relationship between an amplitude B–B′ of vibration of the apex of the probe 1 in the horizontal direction and a measurement target region A–A′ of the sample 3. FIG. 2A shows a relative position of the apex of the probe 1 and the sample 3. FIG. 2B shows a status in which the apex of the probe 1 moves sinusoidally with respect to the time axis. In the probe control apparatus of the present invention, since the gap 4 is controlled by using the atomic force between the probe 1 and the sample 3, it is necessary that the shear force caused by the atomic force acts on the probe 1 sufficiently. Therefore, it is difficult to lessen the horizontal amplitude B–B′ of the apex of the probe 1 to a value less than several tens of nanometers. Therefore, spatial resolution in measurement is limited by the amplitude B–B′. In this embodiment, to avoid such limitation, the current applied to the sample 3 from the probe 1 is applied like a pulse in synchronization with the phase of vibration of the apex of the probe 1. The timing for applying current can be synchronized with any phase. For example, in this embodiment, a pulse voltage is applied from the probe bias power source 14 while the apex of the probe 1 is located in the measurement target region A–A′ which is near the center O of the amplitude. Accordingly, even when the amplitude B–B′ of the apex of the probe 1 in the horizontal direction necessary for controlling the gap 4 is large, the spatial resolution of measurement by tunneling current and luminescence of the tunneling current can be intensified according to smallness of the measurement target region A–A′. As mentioned above, according to the probe control apparatus of the present invention, the probe control apparatus is applicable to a sample in which a nonconductive region or a highly resistive region is mixed with a conductive region, and a conductive transparent probe can be used without using a laser beam, which lowers the SN ratio when measuring weak luminescence caused by tunneling current. In addition, a voltage can be applied between the apex of the probe and the sample for applying a current from the apex of the probe to the sample to cause luminescence. Since the voltage applying mechanism applies a pulse voltage between the apex of the probe and the sample in synchronization with vibration of the probe, spatial resonance for measurement by using tunneling current and luminescence of the tunneling current can be intensified even if the amplitude of the apex of the probe in the horizontal direction necessary for controlling the gap is large. (Conductive Transparent Probe) Next, a conductive transparent probe applicable for use in the above-mentioned probe control apparatus will be described. In order to perform gap control stably by the shear force gap control by using the above-mentioned probe control apparatus, it is necessary for the probe to have a smooth frequency-to-amplitude characteristic (represented by a curve indicating a relationship between frequency and amplitude) with few parasitic vibrations. For realizing this characteristic, it is necessary that the probe and the vibrator be integrated while vibrating, so that the probe vibrates by following faithfully the vibration of the vibrator. In order that the probe and the vibrator vibrate together, it is necessary that the probe has a moderate rigidity for keeping moderate contacting pressure between the probe and the vibrator, and that the probe has a moderate elasticity to vibrate stably. If the probe is so soft that contacting pressure between the probe and the vibrator is small, the probe vibrated by the vibrator jumps (amplitude of the probe exceeds that of the vibrator) from the vibrator, so that the probe does not vibrate together with the vibrator and does not follow faithfully the vibration of the vibrator, and parasitic vibration occurs. Thus, movement of the probe becomes unstable. If the probe is so rigid that the contact pressure is too large, the probe may be broken, or the probe cannot be vibrated at the desired amplitude, so that proper movement cannot be obtained. However, a conventional probe used for STM is short, and a thick metal plating is applied on the surface of the probe for preventing mechanical vibration that may cause noise. Thus, rigidity of the probe is large, so that rigidity and elasticity are not proper for realizing shear force gap control. Therefore, the probe cannot be used for shear force gap control. Therefore, a conductive transparent probe is used as follows in the present invention. FIG. 3 shows a horizontal section of the first embodiment of the conductive transparent probe of the present invention. The conductive transparent probe is made of an optical fiber 21 including a core 22 and a cladding 23. A taper part 25 is provided in the optical fiber 21, wherein the taper part 25 ranges within several hundred micrometers from one end opposed to a sample 32 in the optical fiber 21, and the apex of the taper part 25 is sharpened to a size of the nanometer order. To provide conductivity and a luminescence collection function to the taper part 25 of the nonconductive optical fiber 21, a conductive transparent film 24 having conductivity and transparency is applied on the surface of the taper part 25. In addition, in order to provide conductivity to the optical fiber 21, a first metal film 26 having conductivity is applied on the outer surface of the optical fiber 21. The conductive transparent film 24 and the first metal film 26 are connected electrically. Tunneling current is applied from the apex of the conductive transparent probe to the sample 32, and luminescence caused by the tunneling current is collected from the apex of the same conductive transparent probe. A part ranging from a point apart from the one end by a distance D to the other end of the optical fiber 21 is held by a conductive holding plate 29 formed by a metal plate, for example, so that the conductive transparent probe is mounted on the supporting structure 31 (refer to FIG. 4). The conductive holding plate 29 is connected to a bias power source (not shown in the figure). Since the holding plate 29 contacts the first metal film 26 of the conductive transparent probe electrically, it has a function to provide a current to the conductive transparent probe. The current from the bias power source is supplied to the apex of the conductive transparent probe via the first metal film 26 on the surface of the optical fiber 21 and the conductive transparent film 24 on the taper part 25. The conductive transparent probe is vibrated in a direction perpendicular to the axis of the optical fiber 21 by using a sensor 28 that pushes a point (vibration point) apart from the one end (apex) by a distance d (d<D, about several millimeters). The sensor 28 is attached to a vibrator 30 (FIG. 4), and has a function to transmit vibration of the vibrator 30 to the conductive transparent probe and a function to detect vibration and amplitude of the conductive transparent probe. The sensor 28 is electrically insulated from the conductive transparent probe. When the vibrator 30 operates, the conductive transparent probe vibrates at a frequency and an amplitude corresponding to those of a cantilever of a length D. In order to transmit vibration of the vibrator 30 to the apex of the conductive transparent probe faithfully, it is desirable to shorten the distance d. In addition, in order for the conductive transparent probe to vibrate easily, it is desirable to set the vibration point apart from the part where the conductive transparent probe is held, and to make the distance D as large as possible. According to an experiment, stable operation was obtained and the probe was easy to handle when the distance D was no less than 5 mm and the distance d was about 2–3 mm, wherein the distance D is almost the same as the length of the first metal film 26 from a part adjacent to the taper part 25 to the other end. Next, operation of a shear force gap control system in which the conductive transparent probe of the present invention is implemented will be described. FIG. 4 shows a basic structure of the shear force gap control system in which the conductive transparent probe of the present invention is implemented. Although the shear force gap control system is similar to the probe control apparatus described by using FIG. 1, the structure is simplified in the following embodiments since the conductive transparent probe is mainly described. As shown in the figure, a part of the other end side of the conductive transparent probe is held by the conductive holding plate 29. At this time, the conductive transparent probe is placed on the back side of the sensor 28. Next, the conductive transparent probe is bent a little, and the conductive transparent probe is put on the sensor 28 such that a point a distance d apart from the apex of the conductive transparent probe is placed on the front of the sensor 28. By elastic force caused by the bending, the conductive transparent probe and the sensor 28 contact each other with moderate contacting pressure. Next, the conductive transparent probe is vibrated in a direction perpendicular to the axis of the optical fiber 21 (that is, parallel to the surface of the sample) at a specific frequency. The vibrated conductive transparent probe vibrates as a cantilever having a fixed end that is the part attached to the conductive holding plate 29. The sensor 28 outputs a voltage corresponding to the amplitude of vibration of the conductive transparent probe. While the conductive transparent probe is vibrated at a frequency slightly different from an Eigen frequency, when the frequency is changed slightly, the amplitude of the conductive transparent probe changes greatly. Thus, the conductive transparent probe is vibrated at a frequency slightly different from the Eigen frequency, and the amplitude is monitored by the sensor 28. When atomic force between the conductive transparent probe and the sample 32 becomes large as the conductive transparent probe approaches the sample 32, the atomic force acts on the conductive transparent probe as a shear force in a direction perpendicular to the axis of the optical fiber 21. The shear force acts as a resistance force against vibration of the conductive transparent probe vibrating as a cantilever. Thus, the frequency of the conductive transparent probe changes slightly so that the amplitude of the conductive transparent probe at a monitored frequency is changed. This change of the amplitude is detected as a change of output voltage of the sensor 28. When the amplitude of the output voltage becomes a predetermined value, that is, when the shear force becomes a predetermined value, the approach of the conductive transparent probe to the sample 32 is stopped. After that, the gap between the conductive transparent probe and the sample 32 is controlled such that the amplitude of the conductive transparent probe at the monitored frequency is constant (that is, such that shear force is constant) while performing measurement. Accordingly, stable operation of an AFM (Atomic Force Microscope) can be obtained, wherein the AFM is a microscope performing the gap control by using atomic force (Yang et al. “Near-field differential scanning optical microscope with atomic force regulation”, Appl. Phys. Lett., 60(24), Jun. 15, 1992, for example, can be referred to for more information on conventional AFM). Since the shear force gap control is stably performed irrespective of conductivity of the sample 32, it becomes possible to realize a TL apparatus using the conductive transparent probe, that is applicable to a sample in which a nonconductive region or a highly resistive region is mixed with a conductive region. For satisfying a contacting condition between the conductive transparent probe and the sensor 28 necessary for conducting stable shear force gap control, thickness and length of the first metal film 26 applied on the outer surface of the optical fiber 21 are adjusted, so that flexural rigidity of the conductive transparent probe is adjusted. As a material of the first metal film 26, nickel, stainless steel and the like can be used, for example. However, any other material can be used as long as adhesive force between the material and the surface of the optical fiber 21 is strong and conductivity is high. If the distance D from the end of the conductive transparent probe to the conductive holding plate 29 is equal to or less than several millimeters, the flexural rigidity of the optical fiber 21 becomes large. Therefore, there occurs a case in which the sensor 28 slides on the surface of the conductive transparent probe, so that vibrations of the vibrator 30 do not transfers to the optical fiber 21 faithfully. When the thickness of the first metal film 26 of the outer surface of the optical fiber 21 is smaller than about 0.2 μm, the optical fiber 21 is easily broken by a slight shear force. In addition, if first metal film 26 is thin, electrical resistance from the holding plate 29 to the apex of the optical fiber 21 becomes large, so that it becomes difficult to supply a current to the apex of the optical fiber 21. Therefore, a conductive transparent probe having a thin first metal film 26 is not practical. On the other hand, if the thickness of the first metal film 26 is greater than 10 μm, rigidity of the conductive transparent probe becomes large, so that a large force is necessary for bending the conductive transparent probe. When bending the conductive transparent probe forcibly, plastic deformation occurs so that the shape does not return to its original shape. Therefore, the conductive transparent probe having a thick first metal film 26 is not applicable to the shear force gap control. According to an experiment, when thickness of the first metal film 26 was 0.2–10 μm, the conductive transparent probe 21 had elasticity proper for shear force gap control, and good electrical conductivity, so that the conductive transparent probe had good characteristics for shear force gap control. As mentioned above, when the length of the first metal film 26 from the part adjacent to the taper part 25 to the other end is equal to or greater than 5 mm, and thickness of the first metal film 26 is 0.2–10 μm, contacting pressure between the conductive transparent probe and the sensor 28 becomes a proper value, so that the conductive transparent probe vibrated by the sensor 28 does not jump from the sensor 28. The conductive transparent probe integrates with the sensor 28, and follows vibration of the sensor 28 faithfully. The movement of the conductive transparent probe does not become unstable due to parasitic vibration and the like. The conductive transparent probe is not too stiff and contact pressure is not too large. In addition, the conductive transparent probe does not break, and is oscillated at the desired amplitude. Thus, the conductive transparent probe operates properly. Stress concentrates on a boundary part between a part performing bending vibration as a cantilever and a part held by the conductive holding plate 29. Thus, the boundary part is easily broken. In addition, the optical fiber 21 may be distorted by pressure applied to the conductive transparent probe from the holding plate 29 for fixing the conductive transparent probe, so that there is a case that optical characteristics of the conductive transparent probe degrade. In a second embodiment of the present invention shown in FIG. 5, a second metal film 33 is formed on the side of the other end of the conductive transparent probe, such that the optical fiber 21 is not distorted by a pressure applied to the optical fiber 21 from the holding plate 29, wherein the thickness of the second metal film 33 is larger than that of the first metal film 26. As a result of an experiment, it was found that the thickness of the second metal film 33 needed to be no less than 10 μm, and preferably no less than 50 μm. In this embodiment, if the thickness of metal film between the first metal film 26 and the second metal film 33 changes discontinuously, there is a possibility that the conductive transparent probe will be broken since stress concentrates on the part where the thickness changes discontinuously. Therefore, a transitional part 34 where thickness of metal film changes smoothly is provided between the first metal film 26 and the second metal film 33. By adopting such a structure, the conductive transparent probe can be mounted firmly with reliability by the holding plate 29 without degrading optical characteristics of the conductive transparent probe. In addition, a conductive transparent probe that is hard to break by stress concentration can be realized. FIG. 6 is a section view of the third embodiment of the conductive transparent probe of the present invention. As shown in the figure, in this embodiment, the taper part 25 provided in one end part of the conductive transparent probe is covered with a material 35 through which light cannot pass, and a very small hole is provided at the apex of the taper part 25 opposed to the sample 32. By adopting such a structure, it becomes possible to selectively collect only near optical fields in the tunneling current luminescence. In the above-mentioned configuration of the probe control apparatus for controlling the gap by using shear force, vibration of the vibrator is applied to the conductive transparent probe via the sensor contacting the conductive transparent probe, and the sensor detects changes, due to atomic force, of amplitude of the conductive transparent probe. However, the configuration is not limited to this example. There is following another configuration of the gap control apparatus for using shear force. That is, instead of fixing the conductive transparent probe to the holding plate 29, the conductive transparent probe can be fixed directly to the vibrator, and a laser beam is directed to the conductive transparent probe from the side direction of the conductive transparent probe, and, change of amplitude of the conductive transparent probe is detected by measuring the laser beam modulated by vibration of the conductive transparent probe. The conductive transparent probe can be applied to an apparatus for performing shear force gap control by such method using a laser beam. As mentioned above, according to the conductive transparent probe, length and thickness of the first metal film from a part adjacent to the taper part to the other end are set to values such that the conductive transparent probe vibrates while integrating with the vibrator. Therefore, it becomes possible to realize a conductive transparent probe applicable to shear force gap control while tunneling current can be applied and luminescence collection yield is not degraded. Thus, the present invention produces the effect of enabling stable TL measurement for a sample in which a nonconductive region or a highly resistive region is mixed with a conductive region. Especially, a conductive transparent probe having bending rigidity applicable to shear force gap control can be realized by setting the length of the first metal film to be no less than 5 mm, and setting the thickness of the first metal film to be 0.2–10 μm. In addition, in the conductive transparent probe of the present invention, a second metal film is formed on an outer surface of the other end side, wherein thickness of the second metal film is no less than 10 μm, and the first metal film and the second metal film are connected by using a transitional part whose thickness changes continuously. Therefore, the conductive transparent probe can be held firmly with high reliability without degrading optical characteristics, and the conductive transparent probe is not broken even when bending stress concentrates on a part. Further, in the conductive transparent probe, the taper part is covered by a material through which light cannot pass, and a very small hole is provided on the apex of the taper part covered by the material, wherein the diameter of the hole is smaller than a wavelength of a transmission light. Accordingly, only near optical fields can be collected. The present invention is not limited to the specifically disclosed embodiments, and variations and modifications may be made without departing from the scope of the invention. |
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055901689 | summary | BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to an X-ray microscope for obtaining a transmitted X-ray microscopic image of a specimen such as biological specimen by irradiating the specimen with X-rays and exciting radiation rays. 2. Related Art Statement Various studies and developments for X-ray radiation sources and X-ray optical elements have been advanced, and one of their application systems, an X-ray microscope has been proposed. In X-ray microscope, there are provided various imaging optical elements such as Wolter type optical element which is a kind of the grazing incident optical element, a zone plate optical system utilizing diffraction, and a Schwarzschild optical system including two spherical mirrors having multilayer coatings applied thereon. Particularly, a soft X-ray microscope using soft X-rays has been developed to a study of biological substances, because damage to the biological substances can be reduced. That is to say, in the soft X-ray microscope, the biological specimens can be observed with a high resolution without dyeing or staining. Generally, a wavelength range of the soft X-rays extends from 2.ANG. which is the longest wavelength of the hard X-rays to 1000.ANG. which is the shortest wavelength of the vacuum ultraviolet rays, so that the wavelength region of the soft X-rays partially overlaps with a wavelength region of extreme ultraviolet rays. FIG. 1 is a schematic view showing the Wolter optical element, in which X-rays are made incident upon reflecting surfaces illustrated by solid lines at large incident angles (grazing incident) and are reflected thereby due to the total reflection. FIG. 2 is a schematic view depicting the Fresnel zone plate optical element, in which X-rays are reflected by diffraction. FIG. 3 is a schematic view showing the Schwarzschild optical element using two spherical mirrors each having a multilayer coating applied thereon. These X-ray optical elements are well-known in the art and are described in "X-RAY OPTICAL ELEMENTS and THEIR APPLICATIONS", Sadao AOKI, Applied Physics, Vol 56, No. 3, 1987, pp. 342(44)-351(53), so that their detailed explanation is omitted here. Among the soft X-ray wavelength region, soft X-rays within a wavelength region of .lambda.=43.7.ANG. to 23.6.ANG., i.e. a so-called water window region between the K.alpha. absorption edge of carbon and the K.alpha. absorption edge of oxygen are important, because the absorption of the soft X-rays of this region by carbon and nitrogen is large, while that by water composed of oxygen and hydrogen is small. Therefore, by using the soft X-rays of the water window region, it is possible to observe biological specimens mainly composed of proteins (living tissues) with high resolution in water. Due to this fact, research institutions have endeavored to develop optical elements, radiation sources and detectors having high performance for the soft X-rays of the wavelength region of 43.7 to 23.6.ANG.. As stated above, the soft X-rays within the above mentioned wavelength region are suitable for inspecting the biological substances, however it is practically difficult to manufacture the optical elements, radiation sources, detectors and so on having excellent property due to the following reason. Firstly, it is very difficult to manufacture the X-ray multilayer reflecting mirror and filter having superior characteristics for the soft X-rays of the above mentioned wavelength region. That is to say, upon designing the multilayer reflecting mirror having a high reflectance, it is required that two kinds of substances with the largest possible difference between their refractive indices are built up alternately to form a multilayer film. However, the refractive indices of almost all substances for the X-rays are close to unity, and thus it is difficult to choose two kinds of substances with the large difference in the refractive index. Although a proposal has been made for materials of the multilayer coating whose reflectances are expected to be somewhat improved, such as multilayer films Ni/Sc and Ni/Ti having a structure of laminating alternately Ni (nickel) and Sc (scandium) and Ni and Ti (titanium), these materials are liable to be crystallized during the evaporation, and this makes it difficult to deposit a uniform film. Furthermore, when the normal incident mirrors are to be formed by the presently developed technique, a period or pitch of a multilayer coating for the wavelength region of 43.7 to 23.6.ANG. becomes smaller than 20.ANG., so that the fabrication of the thin multilayer film is difficult. Still further, in the wavelength region of 43.7 to 23.6.ANG., the absorption of X-rays in terms of carbon is high and thus it is impossible to utilize organic materials as filters and a choice of filter materials is limited. In the X-ray microscope, it is necessary to provide the multilayer coatings and filters, but the above problems become obstacles upon utilizing the soft X-rays. Even though the above mentioned first problem were solved, there is remained a second problem which will be explained next. This second problem relates to a quality of the transmitted X-ray image of a specimen, particularly the decrease in contrast of the image. That is to say, the absorption of the soft X-rays by a living specimen is determined by a thickness of the specimen, a density of nitrogen contained in the specimen and a wavelength of the X-rays, and therefore when the specimen has a large thickness and a high density of nitrogen, a substantial part of the X-rays is absorbed by the specimen and thus the transmitted X-ray image of the specimen becomes dark. When a thin specimen having a low nitrogen density is observed, almost all incident X-rays are transmitted through the specimen and thus a transmitted X-ray image becomes bright. In both cases, the constant of the transmitted X-ray image is very low. The above mentioned second problem could be solved by adjusting the thickness of the specimen or by adjusting the wavelength of the X-rays within the wavelength region of 43.7 to 23.6.ANG., because the nitrogen density of the specimen could never be artificially adjusted. In the first solution, the thickness of the specimen is adjusted with the aid of a precision machine such as a microtome which requires high operator skill for cutting the specimen to reduce its thickness. Further, the cutting operation has to be repeated through the rule of trial and error and requires a long time. Therefore, this solution is not practical at all. The second solution requires a wide change in design and layout of the microscope optical systems in using the X-ray optical elements such as zone plate and Schwarzschild optical element, so that this solution is also of little practical use and at variance with the reality. There has been proposed an X-ray microscope using X-rays of a wavelength region of, e.g. 65 to 43.7.ANG. other than the above mentioned region of 43.7 to 23.6.ANG. in which a microscopic image of a specimen of a particular protein molecule can be obtained with high contrast. Now a principle of this X-ray microscope will be explained with reference to FIGS. 4 and 5. FIGS. 4A to 4F represent the transition process of electron in carbon atom upon absorbing X-rays. FIG. 4A shows an electron arrangement within the carbon atom in the ground state. When the carbon atom is irradiated with X-rays, an electron E in the 1s orbit is ionized as illustrated in FIG. 4B (this is referred to as a first transition) and a hole is formed in the 1s orbit as depicted in FIG. 4C. This condition is very unstable in the view point of energy, so that an electron in the 2p orbit is transfers into the 1s orbit (this is termed as a second transition) to secure its stability as shown in FIG. 4D. When the carbon atom constitutes a protein molecule, a hole formed in the 2p orbit (see FIG. 4E) captures an electron (a third transition) from a surrounding constituent element to resume the initial ground state as shown in FIG. 4F. During the above mentioned transition process, the transmitted X-ray microscopic image of protein is obtained by utilizing the first transition. However, if a wavelength of the used X-rays is longer than the absorption edge of the carbon, the X-rays could not be absorbed by the protein, and thus the contrast of the obtained microscopic image is decreased extremely. Now considering the preceding electron transitions from their reverse processes, it is recognized that even though the wavelength of the X-rays is longer than the absorption edge of carbon, the transmitted X-ray microscopic image of protein can be observed with high contrast. At first, from the ground state shown in FIG. 5A, an electron E in the 2p orbit is excited or ionized due to an reversed third transition to form a hole in the 2p orbit as shown in FIG. 5B. Then, as illustrated in FIG. 5C, an electron in the 1s orbit is excited by the irradiation of X-rays into the 2p orbit due to the reversed second transition as depicted in FIG. 5D. This reversed second transition can be performed by the X-rays having a photon energy which is lower than the wavelength of the absorption edge of carbon. That is to say, the reversed second transition can be carried out by the X-rays having a wavelength longer than the absorption edge of carbon. The condition of FIG. 5D is entirely identical with the condition of FIG. 4B which is obtained after the first transition for ionizing the electron in the inner-shell 1s from the ground state, but an energy for ionizing or exciting the electron from the 1s orbit into the 2p orbit is about several to twenty eV (corresponding to a wavelength region of 100 to 300 nm), so that the reversed second transition may be performed by means of an ultraviolet laser. An energy required for exciting the electron from the 1s orbit to the 2s orbit in FIG. 5D is smaller than an energy required for ionizing the inner-shell electron in FIG. 4B by several to 20 eV. Therefore, by using the two step transition including the process for exciting the electron in the 2p orbit and the process for exciting the electron in the 1s orbit into the 2p orbit as shown in FIGS. 5A to 5D, it is possible to the observe the transmitted image of protein even by using the X-rays of the wavelength longer than the absorption edge of carbon. The superiority of the above mentioned method using the reversed transitions has been quantitatively confirmed by J. K. klems in X-ray Absorption in Valence-excited Molecules as a Possible Contrast Mechanism for Chemically Sensitive Imaging and Spectroscopy, Physical Review A, Vol. 43, No. 4, Feb. 1991, pp. 2041-2045. In this method, firstly the X-rays having a wavelength longer than the absorption edge of carbon can be used, and therefore the multilayer coating may be formed by materials such as W (tungsten) and C (carbon) which are excellent in optical constant and easy of film fabrication. Moreover, these materials have been studied for a long time and have been actually used. Secondly, a necessary energy for ionizing or exciting the electron in the 2p orbit differs for particular proteins, so that carbon atom in a specific protein can be selectively excited or ionized. Further, a value of energy for the succeeding electron transition from the 1s orbit into the 2p orbit is determined uniquely. Therefore, when X-rays having the equivalent photon energy are taken as a probe, it is possible to obtain the transmitted X-ray image of a desired protein. In this case, the contrast of this transmitted X-ray microscopic image is enhanced by more than one figure compared with the conventional method utilizing the wavelength region of 43.7.ANG. to 23.6.ANG. as shown in FIG. 6. The above mentioned principle can be easily realized by slightly changing the existing X-ray microscope system. FIG. 7 is a schematic view showing the known X-ray microscope. The X-ray microscope comprises an X-ray source 1 for emitting X-rays having a given wavelength, a condenser lens 2 for projecting the X-rays onto a specimen 3, an objective lens 4, a filter 5 and a detector 6 which are arranged on the same optical axis. The objective lens 4 may be classified into two groups, i.e. a wave dispersion type such as zone plate or the Schwarzschild optical element and a grazing incident mirror type of collecting white light such as Wolter type optical element. When a white light source is used for the X-ray source 1 and the Wolter type objective lens 4 is provided, it is necessary to arrange a spectrometer on the optical path extending to the detector 6. The X-ray detector 6 may be formed by a microchannel plate (MCP) and an imaging element such as charge coupled device (CCD). When the white light radiation source is used, a thin film filter such as beryllium (Be) film for cutting off stray light rays having wavelengths longer than that of ultraviolet is generally arranged in the optical path. In order to avoid the absorption of the X-rays by the air, the above mentioned optical elements are all arranged within a vacuum chamber not shown. The X-ray detector 6 is connected to a signal processing circuit and an image signal produced by this circuit is supplied to a monitor to display a visible image of the specimen on the monitor. The reversed transition method proposed by J. H. Klems has been applied to the above mentioned X-ray microscope by simply adding ultraviolet ray source 7, condenser lens 8 and ultraviolet (UV) reflection mirror 9 as illustrated in FIG. 8. It should be noted that in U.S. Pat. No. 5,216,699 issued on Jun. 1, 1993 and assigned to the same assignee to whom the present application is also assigned, there is described the X-ray microscope shown in FIG. 8. The UV reflection mirror 9 is inserted between the specimen 3 and the objective lens 4 and has a sufficiently high transmittance for the wavelength region of 65 to 43.7.ANG. and has a sufficiently high reflectance for the ultraviolet rays. Therefore, the UV reflection mirror 9 also serves as the X-ray filter for cutting off the noise, i.e. stray light rays having wavelengths longer than that of the ultraviolet, so that the X-ray filter 5 shown in FIG. 7 is dispensed with. In this X-ray microscope, when the specimen 3 is irradiated with the X-rays having the equivalent energy for effecting the reversed second transition shown in FIG. 4p as well as the ultraviolet rays emitted from the UV source 7 by means of the condenser lens 8 and UV reflection mirror 9, the electron in the 2p orbit of carbon of a specific protein can be ionized or excited and the transmitted X-ray microscopic image of the specimen 3 can be observed through the process of the transition proposed by J. H. Klems. The X-ray microscope shown in FIG. 8 has further advantages. That is to say, the absorption coefficient of the X-rays due to a living specimen can be simply changed, so that the contrast of the transmitted X-ray image can be adjusted without changing the thickness of the specimen or the wavelength of the X-rays. The X-ray absorption coefficient of the living substance is proportional to the number of carbon atoms having the holes in the 2p orbits after the irradiation with the ultraviolet rays as shown in FIG. 5C and this number is proportional to a photon flux or an amount of the irradiated ultraviolet rays. Therefore, by adjusting a photon flux or an amount of the ultraviolet rays to be made incident upon the living specimen, the absorption coefficient of the X-rays in terms of the living specimen can be changed such that the contrast of the transmitted X-ray image becomes optimum. Moreover, a transmitted X-ray image of the specimen without the irradiation with the ultraviolet rays is picked-up in addition to the transmitted X-ray image of the same specimen with the irradiation with the ultraviolet rays, and then a differential image of these two X-ray images is derived to remove background noise due to elements other than carbon. The thus obtained differential X-ray image has a superior contrast purely due to carbon. FIGS. 9 and 10 are schematic diagrams showing another embodiments of the X-ray microscope disclosed in the above mentioned U.S. Pat. No. 5,216,699. The X-ray microscope shown in FIG. 9 comprises an X-ray source 11 formed by a synchrotron radiation (SOR) source, a spectrometer 12, a condenser lens 13 formed by a Fresnel zone plate, an objective lens 14 also formed by a Fresnel lens, and an X-ray detector 16 formed by MCP. These elements are arranged on the same optical axis. A specimen 14 to be inspected is placed between the condenser lens 13 and the objective lens 15. The X-ray microscope further comprises an ultraviolet laser light source 17, a condenser lens 18, a glass wedge 19 and a thin film 20 made of Be (beryllium) which is arranged at 45 degrees with respect to the optical axis. Ultraviolet rays emitted by the UV light source 17 is projected onto the specimen 14 by means of the Be thin film 20. This Be thin film 20 further serves to prevent stray rays such as ultraviolet rays from being incident upon the detector 16. The wedge 19 is made of a material such as BK7 glass having a high absorption for the ultraviolet rays and is arranged movably with respect to an optical axis as shown by a double headed arrow, so that by adjusting a position of the wedge 19, an optical path length of the wedge through which the ultraviolet rays pass can be changed so as to adjust an amount of ultraviolet rays to be made incident upon the specimen 14. In this manner, a properly adjusted amount of the ultraviolet rays can be projected onto the specimen 14 together with the X-rays, and thus it is possible to obtain a transmitted X-ray microscopic image of the protein specimen having a high contrast compared with the known X-ray microscope. The X-ray microscope illustrated in FIG. 10 is basically same as that shown in FIG. 9. In this microscope, there is provided a laser plasma source including Nd:YAG laser 21, an X-ray condenser lens 22 is formed by a Wolter type optical element, an objective lens 23 is formed by a Schwarzschild optical element, and a detector 24 is formed by a microchannel plate (MCP). In order to generate X-rays, a laser light beam emitted from the Nd:YAG laser 21 is made incident upon a target 25 by means of a half mirror 26 and a condenser lens 27. A part of the laser beam reflected by the half mirror 26 is transmitted through a polarizer 28 to adjust an amount of laser beam passing therethrough, and then the laser beam emanating from the polarizer is made incident upon an optically anisotropic or non-linear crystal 29 such as KDP (KH.sub.2 PO.sub.4). Then, the ultraviolet rays are converted into fourth order harmonics and the thus converted ultraviolet rays of harmonics are made incident upon an ultraviolet reflecting mirror 32 by means of reflection mirror 30, condenser lens 31 and UV. The ultraviolet rays reflected by the mirror 32 are then made incident upon a specimen 33. The X-ray microscope shown in FIG. 10 has advantages compared with the X-ray microscope depicted in FIG. 9 that there is not provided the UV light source such as UV laser and the objective lens 23 is formed by the Schwarzschild optical element including the multilayer coatings made of W/C and having an excellent optical property. For instance, the reflectance of a multilayer film composed of 200 membranes coating for the normal incident ultraviolet having the wavelength of 45.ANG. amounts to about 30%, so that it is possible to observe a transmitted X-ray image having a high brightness. In the X-ray microscopes illustrated in FIGS. 8 to 10, it is essentially required to irradiate the specimen with a sufficiently large amount of the ultraviolet rays in order to observe a transmitted X-ray microscopic image having a high contrast. However, in these X-ray microscopes, the UV light sources are arranged outside the vacuum chamber and the ultraviolet rays are made incident upon the specimen by means of the UV transmissive window provided in the wall of the vacuum chamber, so that a relatively large amount of the ultraviolet rays is absorbed by the air and window and thus an amount of the ultraviolet rays actually impinging upon the specimen is reduced. Therefore, the transmitted X-ray microscopic image having the high contrast could not be obtained. In order to avoid such a drawback, it is necessary to provide a large scale UV light source which can emit a very large amount of ultraviolet rays. But this solution results in high cost. Furthermore, in the above mentioned X-ray microscopes, the ultraviolet rays are made incident upon the specimen as a diverging ultraviolet beam by means of the thin BE film serving as the ultraviolet reflecting mirror and X-ray filter. Therefore, a photon flux measured at a surface of the specimen is liable to be small, so that the electron in the 2p orbit could not be effectively ionized or excited and the contrast of the transmitted X-ray image is liable to be decreased. Moreover, a part of the ultraviolet rays is scattered within the vacuum chamber and stray rays are made incident upon the X-ray detector. This results in white noise in the transmitted X-ray microscopic image and deteriorates the image quality. As stated above, in the X-ray microscopes shown in FIGS. 8 to 10, the X-ray absorption coefficient of the specimen can be changed by adjusting an amount of the ultraviolet rays impinging upon the specimen without changing a thickness of the specimen. However, the inventors of the present application have found that it is practically difficult to observe the transmitted X-ray microscopic image having a high contrast without adjusting a thickness of the specimen due to the existence of elements contained in a portion of the specimen which portion is free from the irradiation of the ultraviolet rays. In order to observe an optimum X-ray image having a good contrast, the inventors have found that a mutual relationship between a thickness of a specimen, a wavelength of the X-rays and a resolving power of tone has to be established quantitatively. However, no one has proposed such a relationship. In the X-ray microscopes mentioned above, in order to obtain a transmitted X-ray microscopic image having a good contrast, it is necessary to determine a photon flux or an amount of the ultraviolet rays to be made incident upon the specimen. In other words, the inventors have found that the photon flux of the ultraviolet rays is one of important parameters for observing the X-ray image having an excellent quality. Moreover, if a suitable photon flux is determined, it will be possible to select a suitable ultraviolet light source. The selection of the ultraviolet light source such as a laser is a very important factor for designing and manufacturing actual products and puts a large influence upon cost and performance of products. However, there has not been established a theory for determining the photon flux of the ultraviolet rays to be made incident upon the specimen. The inventors have further found experimently that the quality of the transmitted X-ray microscopic image obtained by the ultraviolet excitation type X-ray microscope depends on a time period during which the specimen is irradiated with the ultraviolet rays and a timing of the irradiation of the X-rays with respect to the irradiation of the ultraviolet rays. Also in this case, the photon flux of the ultraviolet to be made incident upon the specimen is an important factor. In the ultraviolet excitation type X-ray microscope explained above, in order to observe various kinds of elements contained in a specimen or in order to observe the same element contained in different substances of the specimen, it is necessary to change a wavelength of the ultraviolet rays. However, in the above mentioned X-ray microscopes, a wavelength of the ultraviolet rays could not be changed or adjusted in accordance with objects to be observed. As the soft X-ray prove, there has been proposed a secondary electron spectroscopic apparatus, in which particle beam such as electron beam, proton beam, positron beam, neutron beam, and photon beam is projected onto a specimen to emit secondary electrons and a power spectrum of the secondary electrons is detected. Recently, in a field of analyses for semiconductor surface, carbon containing organic substances, semiconductor process such as CVD, and organic electronic devices, there has been required to develop a new estimation by utilizing the electron spectroscopy for chemical analysis (ESCA) and the Auger electron spectroscopy, in which soft X-rays having wavelengths longer than several .ANG. are used as an optical probe. Particularly, there has been required to develop an analysis using soft X-rays having a wavelength longer than 5--as the optical probe for investigating biological substances including oxygen (K absorption edge is 23.32.ANG.), nitrogen (K absorption edge is 30.99.ANG.), carbon (K absorption edge is 43.68.ANG.), phosphorus (L absorption edge is 94.ANG. and K absorption edge is 5.8.ANG.), calcium (L absorption edge is 35.ANG.), sodium (K absorption edge is 11.6.ANG.), magnesium (K absorption edge is 9.5.ANG.). In presently available estimating apparatuses, a radiation source is formed by an X-ray tube, and thus use may be made of characteristic X-rays having a wavelength shorter than several .ANG.. Therefore, when a specimen mainly composed of carbon is to be estimated, its absorption coefficient is too small to yield a large amount of photoelectrons or Auger electrons, so that a sensitivity of analysis is liable to be low. Further, only the characteristic X-rays can be used, it is impossible to perform various analyses and elements can not be judged precisely. In view of the above fact, there has been desired to develop a novel estimation using the soft X-rays having a wavelength longer than several .ANG.. However, in order to obtain white soft X-rays, it is necessary to provide a large scale synchrotron radiation source (SOR) which could be hardly utilized by general users. In order to avoid the above mentioned drawbacks, there has been proposed in Japanese Patent Laid-open Publication Kokai Hei 4-140651 an electron spectroscopic analyzing apparatus using a laser plasma light source. In this analyzing apparatus, a laser beam having an intensity higher than 10.sup.12 W/cm.sup.2 is projected upon a target made of a metal under a pressure lower than 10.sup.-4 Torr and the target metal is brought into a plasma condition to emit soft X-rays having a wavelength longer than 5.ANG.. Therefore, the light source can be simply realized by means of easily available YAG laser and vacuum chamber. The soft X-rays emitted by the above mentioned laser plasma can be advantageously dispersed widely by a toroidal grating monochrometer rather than by a constant-deviation monochrometer. Further, by providing a slit on a Rowland circle of the toroidal grating monochrometer, it is possible to derive soft X-rays having a given wavelength. The thus obtained soft X-rays are then made incident upon a specimen to emit secondary electrons. The secondary electrons are then detected by an electron analyzer arranged at a given angle with respect to the specimen and the energy of secondary electrons is analyzed. In this manner, elements constituting a surface of the specimen can be judged or determined precisely. That is to say, by analyzing the energy of electrons ionized by photons or Auger electrons, it can be determined how much electrons are emitted from what energy levels of what elements. Further by selecting a wavelength of the soft X-rays, an amount of Auger electrons emitted from a specific element can be exclusively increased. For instance, when X-rays having a wavelength near the K absorption edge of carbon is projected onto the specimen, carbon KLL Auger electrons having the kinetic energy of about 250 eV can be predominantly observed, and thus an amount of carbon contained in a specimen surface can be analyzed. By using the above mentioned electron spectroscopic analyzing apparatus, it is possible to observe Auger electrons emitted by various elements by selecting a wavelength of the X-rays, so that the analysis for elements can be performed with a very high sensitivity compared with the other type ESCA using the X-ray tube. Moreover, by scanning the wavelength of the X-rays and detecting amounts of emitted secondary electrons, it is possible to effect an analysis utilizing the extended X-ray absorption fine structure (EXAFS). Further, when an X-ray optical system such as inclined incident mirror is arranged behind the slit provided on the Rowland circle to produce an X-ray microbeam and a specimen stage is scanned with the X-ray microbeam, it is possible to obtain a two-dimensional image representing a distribution of an element under inspection. In this case, if the X-ray optical elements are formed by the Schwarzschild optical element or zone plate having a wavelength dependent dispersion, the monochrometer may be dispensed with. However, when a specimen is composed of a plurality of substances and these substances have the same element, it is impossible to observe the element contained in a particular substance. For instance, a biological specimen contains various proteins and these proteins contain carbon, so that when the specimen is irradiated with the X-rays, every carbon elements contained in all the proteins emit the secondary electrons. Therefore, it is impossible to derive a distribution of the carbon element within a particular protein. SUMMARY OF THE INVENTION The present invention has for its general object to provide an X-ray microscope, in which a transmitted X-ray microscopic image of a specimen having an excellent image quality can be observed. It is another object to provide an X-ray microscope, in which a transmitted X-ray microscopic image of a specimen can be observed with a high contrast without decreasing a photon flux of ultraviolet rays. It is another object of the invention to provide an X-ray microscope, in which a specimen can be irradiated with ultraviolet rays under a desired condition, so that a transmitted X-ray microscopic image of a specimen can be observed with a desired contrast. It is another object of the invention to provide an X-ray microscope, in which a transmitted X-ray microscopic image of a specimen can be observed by suitably determining a mutual relationship between a thickness of the specimen, a wavelength of X-rays and a tone resolving power of the X-ray image. It is another object of the invention to provide an X-ray microscope, in which a specimen can be irradiated with ultraviolet rays under a suitable photon flux and a transmitted X-ray microscopic image of the specimen can be observed with a desired contrast. It is another object of the invention to provide an X-ray microscope, in which a specimen can be irradiated with ultraviolet rays and X-rays at suitable timings to observe an X-ray microscopic image with a desired contrast. It is another object of the invention to provide an X-ray microscope, in which a wavelength of ultraviolet rays can be adjusted in accordance with substances under inspection. It is still another object of the invention to provide a secondary electron spectroscopic apparatus, in which it is possible to observe selectively a particular element contained in a substance which constitutes a specimen together with other substances which contain the same element. According to a first aspect of the present invention, in an X-ray microscope in which a specimen is irradiated with X-rays having a wave length region of 65 to 43.7.ANG. and ultraviolet rays and X-rays transmitted through the specimen are received by an X-ray detector to form a transmitted X-ray microscopic image of the specimen, the improvement being characterized in that a non-linear optical medium is provided in a vacuum chamber in which an X-ray optical system of the X-ray microscope is installed, radiation rays having a wavelength longer than that of the ultraviolet rays are made incident upon the non-linear optical medium to convert said radiation rays into ultraviolet rays, and the thus converted ultraviolet rays are made incident upon the specimen. According to a second aspect of the invention, in an X-ray microscope in which a specimen is irradiated with X-rays having a wave length region of 65 to 43.7.ANG. and ultraviolet rays and X-rays transmitted through the specimen are received by an X-ray detector to form a transmitted X-ray microscopic image of the specimen, the improvement being characterized in that a ultraviolet transmissive window is provided in a wall of a vacuum chamber in which an X-ray optical system of the X-ray microscope is arranged and the ultraviolet rays are made incident upon the specimen through said window as a converged or parallel ultraviolet beam. According to a third aspect of the invention, in an X-ray microscope for forming a transmitted X-ray microscopic image of a specimen by irradiating the specimen with soft X-rays and ultraviolet rays, the improvement being characterized in that the X-ray microscope is constructed to satisfy the following condition; EQU Z<Log.sub.e M/(2r.sub.e .lambda.N.sub.0 f) wherein Z: thickness of specimen PA0 r.sub.e : classical electron radius PA0 .lambda.: wavelength of X-ray PA0 N.sub.0 : the number of molecules or atoms under observation in unit volume PA0 f: imaginary part of atomic scattering factor at .lambda. PA0 M: resolving power of tone of image. PA0 I.sub.0 : the number of photons irradiating specimen per unit time per unit area (photon flux) PA0 .tau.: lifetime of molecules or atoms excited by irradiation with exciting radiation rays PA0 .sigma..sub.UV : excitation cross-section of molecules or atoms under observation due to exciting radiation rays PA0 .sigma..sub.X : cross-section of X-rays for exciting outer-shell electron of molecule or atom under observation into excitation-generated outer-hole PA0 M: tone resolving power of image PA0 .tau.: lifetime of molecule or atom under observation excited with the exciting radiation rays PA0 T: time period of irradiation with the exciting radiation rays. According to a fourth aspect of the present invention, an X-ray microscope for forming a transmitted X-ray microscopic image of a specimen by irradiating the specimen with soft X-rays and exciting radiation rays, the improvement being characterized in that the exciting radiation rays having a photon flux which satisfies the following condition is made incident upon the specimen; EQU [exp{.sigma..sub.UV /(.sigma..sub.X M)}-1l]/(.sigma..sub.UV .tau.)<I.sub.0 wherein According to a fifth aspect of the invention, in an X-ray microscope for forming a transmitted X-ray microscopic image of a specimen by irradiating the specimen with soft X-rays and exciting radiation rays, the improvement being characterized in that after an initiation of irradiation with the exciting radiation rays, irradiation with said soft X-rays is started within a time period of (T+3.tau.); wherein According to a sixth aspect of the invention, in an X-ray microscope for forming a transmitted X-ray microscopic image of a specimen by irradiating the specimen with soft X-rays and exciting radiation rays, the improvement being characterized in that a wavelength of said exciting radiation rays is changed in accordance with a substance under observation contained in the specimen. According to a seventh aspect of the present invention, a secondary electron spectrometer comprising: an X-ray radiation source for emitting X-rays; an exciting radiation source for emitting exciting . radiation rays; |
description | This application is a continuation of prior International Patent Application No. PCT/JP2007/061654, filed Jun. 8, 2007, the entire contents of which are incorporated herein by reference. 1. Field of the Invention The present invention relates to: a charged particle beam inspection apparatus used to observe or inspect a sample while emitting a charged particle beam such as an electron beam onto the sample; and an inspection method using a charged particle beam. 2. Description of the Prior Art In a manufacturing process of a semiconductor device, an electron beam apparatus such as an electron microscope is used to observe a sample or to measure a line width of a pattern or the like. During the observation or measurement of the sample, such an electron beam apparatus scans the sample while emitting an electron beam onto an observation target portion of the sample, and converts the amount of electrons such as secondary electrons into luminance which is then displayed as an image on a display device. During such observation or measurement of the sample, an electron beam is emitted (irradiated) onto the sample. Here, this emission of the electron beam causes a phenomenon in which a surface of the sample is charged. Specifically, the irradiated surface of the sample is charged positively or negatively depending on the difference between the charge of charged particles incident on the sample and the charge of charged particles emitted therefrom. Additionally, in accordance with the charge potential on the surface of the sample, the emitted secondary electrons may be accelerated or drawn onto the sample. This changes efficiency of emitting secondary electrons. Consequently, a problem arises that results obtained from detecting the amounts of electrons cannot be stable. Moreover, if the charge potential is unchanged, it does not adversely affect anything, however, if the charge potential changes with time, the charge potential on the surface of the sample changes an acceleration condition or a deflection condition of a primary electron beam. This causes a problem that an accurate measurement is not maintained. To address these problems, various methods to prevent charging on a sample have been proposed. As a technique related to the above, Japanese Patent Application Publication No. 2003-142019 discloses a method of controlling charging on a sample surface by use of an acceleration voltage at which the yield of the secondary electrons is greater than 1 and an acceleration voltage at which the yield of the secondary electrons is smaller than 1. This method of controlling the charging on the sample surface with use of the acceleration voltages may not ensure the measurement accuracy before and after the changes of the acceleration voltages. During the sample observation using the electron beam apparatus, the phenomenon of charging the sample occurs as described above. In this respect, if electrical connection of the sample is possible, as in the case of a semiconductor wafer, for example, the charging phenomenon on the sample is less likely to occur, because the wafer can discharge electricity by having a conductor on the electrically connected wafer grounded. In this case, charging is not a problem in practical use. On the other hand, if the sample is non-conductive, or if the sample, even though made of a conductive material, is incapable of being grounded and thus is in an electrically floating state, there arises a problem that a charging state of the sample changes with time, so that a measurement target image drifts accordingly. In a case, for example, where dimensions of a photomask used as an original plate for exposing a semiconductor are measured, charging occurs in the following two states. The first state is where a conductor such as chromium entirely lies on a glass substrate which is in the course of having wirings manufactured and where a resist wiring used for etching a wiring to chromium lies on the conductor. The second state is where the processing of manufacturing the wirings are completed so that a wiring made of the conductor such as chromium lies on the glass substrate. Especially in the measurement after the etching process, most of the wiring patterns on the mask become incapable of being electrically grounded and thus are in an electrically floating state. In such a case, the surface potential of the measurement target wiring becomes extremely unstable. This charging phenomenon changes with time and continues until the potential is stabilized. Therefore, after the potential is stabilized, a sample can be observed or measured stably. However, it takes a long time, several tens of seconds, for example, until the potential is finally stabilized. This leads to a problem of deterioration in throughput of measurement processing. The present invention has been made in consideration of the above-described problems in the conventional techniques, and aims to provide: a charged particle beam inspection apparatus which is capable of easily stabilizing the potential on a sample surface in a short time, and is capable of measuring a sample with accuracy; and an inspection method using a charged particle beam. The above-described objects are achieved by a charged particle beam inspection apparatus including: an electron gun emitting an electron beam; a first condenser lens and a second condenser lens used to focus the electron beam; a beam control panel provided between the first condenser lens and the second condenser lens; and a control unit performing stabilizing processing in which excitation currents respectively supplied to the first condenser lens and the second condenser lens are set to havc predetermined values, thereby the current amount of the electron beam passing through an opening of the beam control panel is regulated so that the electron beam to be emitted (irradiated) onto the sample has a larger current amount than that at a measurement, and then the electron beam is emitted onto the sample for a predetermined time period. In the above-described charged particle beam inspection apparatus, after finishing the stabilizing processing, the control unit may set the values of the excitation currents back to values for the measurement, and then the measurement of a dimension of the sample may be performed, the excitation currents respectively supplied to the first condenser lens and the second condenser lens. Moreover, in the above-described charged particle beam inspection apparatus, in the stabilizing processing, the control unit may set the excitation current to be supplied to the first condenser lens lower smaller than the excitation current supplied thereto for the measurement, to increase a focal distance of the first condenser lens, and may set the excitation current to be supplied to the second condenser lens highcr larger than the excitation current supplied thereto for the measurement, to regulate a focal distance of the second condenser lens to be equal to a focal distance thereof for the measurement. In the present invention, the excitation current supplied to the first condenser lens in the stabilizing processing is set to be smaller than that supplied thereto for the measurement, in order to increase the focal distance of the first condenser lens. Consequently, the area of the beam control panel on which the electron beam is emitted is narrowed so that the amount of the electron beam passing through the opening of the beam control panel is relatively increased compared to that for the measurement. As a consequence, the amount of current of the electron beam emitted (irradiated) on the sample is increased. The emission, for a predetermined time period, of a large amount of electron beam thus regulated makes it possible to stabilize the potential on a sample surface in a short time. As a result, a track of the electron beam is made steady, and dimensions of a pattern on the sample can be measured with accuracy. Moreover, the above-described charged particle beam inspection apparatus may further include a storage to store the values of the excitation currents to be respectively supplied to the first condenser lens and the second condenser lens. The control unit may sequentially: extract, from the storage, a value of a first excitation current to be supplied to the first condenser lens, a value of a second excitation current to be supplied to the second condenser lens, and an emission time period, the values and the emission time period set for the stabilizing processing; supply the first excitation current to the first condenser lens to make a focal distance of the first condenser lens longer than a focal distance thereof for the measurement; supply the second excitation current to the second condenser lens to regulate a focal distance of the second condenser lens to be equal to a focal distance thereof for the measurement; and emit the electron beam onto the sample for the emission time period to stabilize a potential on a surface of the sample. In the present invention, the amount of current to be supplied to each of the first and the second condenser lenses (one of stabilizing conditions) is obtained in advance in order to stabilize the above-described potential. Here, the amount of current corresponds to a desired time period taken for stabilizing. The stabilizing conditions for each sample are stored in the storage. Since the charged particle beam inspection apparatus automatically performs the stabilizing processing with reference to the stabilizing conditions stored in the storage, a user can save time and labor considerably. Meanwhile, the above-described problems are solved by a charged particle beam inspection apparatus including: an electron gun which emits an electron beam; a first condenser lens forming, upon receipt of the electron beam emitted from the electron gun, a first focused beam which focuses at a position located a predetermined first focal distance away from the first condenser lens; a beam control panel forming a passed beam upon receipt of the first focused beam, the passed beam formed of a portion of the first focused beam passing through a predetermined opening formed in the beam control panel and located at an axial center; a second condenser lens forming, upon receipt of the passed beam, a second focused beam which focuses at a position located a predetermined second focal distance away from the second condenser lens; an objective lens emitting, upon receipt of the second focused beam, the second focused beam onto an irradiation target sample a predetermined third focal distance away from the objective lens; and a control unit controlling, in order to stabilize a potential on a surface of the sample, focusing conditions for the first condenser lens and the second condenser lens by changing the first focal distance and the second focal distance in a way that the passed beam passing through the opening becomes larger. In addition, another aspect of the present invention is an inspection method using the charged particle beam inspection apparatus according to the above aspect. The inspection method according to the aspect is to measure a dimension of a pattern formed on a sample by use of a charged particle beam inspection apparatus, the apparatus including: an electron gun emitting an electron beam; a first condenser lens and a second condenser lens to focus the electron beam; a beam control panel provided between the first condenser lens and the second condenser lens; a storage to store values of excitation currents to be respectively supplied to the first condenser lens and the second condenser lens; and a control unit for controlling the electron gun, the first condenser lens, and the second condenser lens. The method includes the steps of: extracting, from the storage, the value of the excitation current to be supplied to the first condenser lens, the value of the excitation current to be supplied to the second condenser lens, and an emission time period; supplying the excitation currents respectively to the first condenser lens and the second condenser lens to regulate the electron beam passing through an opening of the beam control panel so that the electron beam to be emitted (irradiated) onto the sample has a larger current amount than that at the measurement; emitting the electron beam onto the sample for the emission time period which is extracted from the storage; and after the emission of the electron beam, and after the foregoing steps for stabilizing processing, measuring the pattern on the sample by setting the values of the excitation currents to be supplied respectively to the first condenser lens and the second condenser lens back to values for the measurement. Hereinbelow, descriptions will be given of an embodiment of the present invention with reference to the drawings. Firstly, a description will be given of a configuration of a charged particle beam inspection apparatus. Secondly, a description will be given of processing to stabilize a potential on a sample surface, which is an aspect of the present invention. Subsequently, a description will be given of an inspection method using a charged particle beam by use of the charged particle beam inspection apparatus. (Configuration of Charged Particle Beam Inspection Apparatus) FIG. 1 is a configuration diagram of the charged particle beam inspection apparatus according to the embodiment. The charged particle beam inspection apparatus 100 mainly includes: an electron scanning unit 10; a signal processing unit 30; a display unit 40; and a control unit 20 for controlling the electron scanning unit 10, the signal processing unit 30, and the display unit 40. As shown, the electron scanning unit 10 includes an electron lens tube unit (column) 15 and a sample chamber 16. The electron lens tube unit 15 includes an electron gun 1, a condenser lens unit 2, a deflection coil 3, and an objective lens 4. The condenser lens unit 2 includes a first condenser lens 2a, a second condenser lens 2b, and a beam control panel 2c having an opening 2d. The beam control panel 2c is disposed between the first condenser lens 2a and the second condenser lens 2b. The sample chamber 16 includes: an XYZ stage 5 which allows a sample 7 to be mounted thereon; and an electron detector 8 which detects secondary electrons or the like emitted from the sample 7. In addition, the sample chamber 16 is connected to a motor (not shown) and an evacuator (not shown). The motor is used to move the XYZ stage 5, and the evacuator is used to keep the inside of the sample chamber 16 at a predetermined low pressure atmosphere. An electron beam 9 emitted from the electron gun 1 is designed to be emitted (irradiated) onto the sample 7 mounted on the XYZ stage 5 through the condenser lens unit 2, the deflection coil 3, and the objective lens 4. The amount of secondary electrons or reflection electrons emitted from the sample 7 in response to the emission (irradiation) of the electron beam 9 is detected by the electron detector 8 which includes a secondary electron control electrode 8a, a scintillator, and the like. An AD converter in the signal processing unit 30 converts the detected amount of secondary electrons into a digital amount, and further converts the digital amount into a luminance signal. Then, the luminance signal is displayed on the display unit 40. An electron deflection amount of the deflection coil 3 and an image scanning amount of the display unit 40 are controlled by the control unit 20. The control unit 20 is formed of a microcomputer and stores a program to execute a measurement. In addition, the control unit 20 determines an acceleration voltage of the electron beam 9, and applies the acceleration voltage to the electron gun 1 to which the control unit 20 is electrically connected. The charged particle beam inspection apparatus 100 configured in the above-described manner performs processing to stabilize the potential on the sample 7 mounted on the XYZ stage 5 in order to observe or measure the sample 7. (Processing to Stabilize Potential on Sample Surface) Processing to stabilize the potential on a sample surface is performed prior to observation or measurement of the sample 7 to make the potential on the surface of the sample 7 uniform. The principle of this processing will be described below. FIG. 2 schematically shows a relationship between the energy of primary electrons and the ratio of secondary electron emission. As shown in FIG. 2, as the energy of the primary electrons is increased from low to high, the ratio of secondary electron emission also increases and reaches 1 when the energy of the primary electron is at an energy level E1. With a further increase of the energy of the primary electrons to reach to an energy level Em, the ratio of secondary electron emission reaches the maximum. When the energy of the primary electrons goes beyond an energy level E2, the ratio of secondary electron emission becomes lower than 1. In this respect, values of the energy of the primary electrons at the energy levels E1, Em, and E2 change depending on a material used in a sample, but where the value at the point Em falls within 500 [eV] to 1000 [eV] in most of the cases. FIGS. 3A and 3B each schematically show a relationship between the ratio of secondary electron emission and a charging state on a surface of an insulating film of a sample 7. FIG. 3A shows a case where the ratio of secondary electron emission is greater than 1. In a range where the ratio of secondary electron emission is greater than 1, the number of the secondary electrons 42 emitted from the sample 7 exceeds the number of the primary electrons 41 incident on the sample 7. Thus, the surface of the sample 7 is positively charged. In contrast, FIG. 3B corresponds to a case where the ratio of secondary electron emission is smaller than 1 and where the energy of the primary electrons is lower than that at the energy level E1 or is higher than that at the energy level E2 shown in FIG. 2. In a range where the ratio of secondary electron emission is smaller than 1, a number of electrons remain on the surface of the sample 7, so that the surface of the sample 7 is negatively charged. When the energy of the primary electrons is sufficiently large and the ratio of secondary electron emission is smaller than 1, the surface of the sample 7 is negatively charged, and thus the primary electrons decelerate in the vicinity of the sample 7. This charging phenomenon proceeds until the energy of the primary electrons is reduced to that at the point energy level E2 and the ratio of the secondary electron emission comes close to 1. The charging voltage in this charging is the difference between the energy at the energy level E2 and the energy of the primary electrons, and the surface of the sample 7 may be charged at a large negative value (α value less than −100 [V], for example) in some cases. Once such a charging occurs, a secondary electron image is largely distorted and a measurement error becomes large. On the other hand, when the ratio of secondary electron emission is greater than 1, the surface of the sample 7 is positively charged. Nevertheless, the surface of the sample 7 when charged by several voltages [V] draws relatively large amount of the secondary electrons 43 back thereto, since the secondary electrons 43 only have several voltages [eV] of energy. An incident current obtained by adding, to the primary electrons, the secondary electrons drawn back to the surface is evenly balanced with an emission current including the emitted secondary electrons. Accordingly the charging does not proceed any further. For this reason, the observation or the like of the sample 7 is made in the range where the ratio of the secondary electron emission is greater than 1. Conventionally, uniform charging has been obtained by controlling changes in potential on a sample surface by changing energy of an emitted electron beam in order that the potential of the sample surface is charged positively or negatively. This control takes time until the potential of the sample surface is stabilized, thereby leading to a problem of deterioration in throughput. The present embodiment has been made focusing on controlling condenser lenses in order to obtain a current needed to stabilize the potential of a sample surface in a short time. FIG. 4 is a schematic diagram showing a track of an electron beam 9 emitted at the measurement (in a regular state). In FIG. 4, a first condenser lens CL1, a second condenser lens CL2, and an objective lens OL are each schematically shown as a convex lens shown by a broken line. The electron beam 9 emitted from the electron gun 1 is converged by the first condenser lens CL1, and then is focused on a point FC11 which is located a distance (focal distance) d11 from the principal surface of the first condenser lens CL1. Here, the principal surface is a surface which is perpendicular to an optical axis and passes through the center of a lens schematically shown. The electron beam 9 having passed through the opening 2d of the beam control panel 2c is further converged by the second condenser lens CL2, and is focused on a point FC2 which is located a distance d12 from the principal surface of the second condenser lens CL2. The electron beam 9 is then converged by the objective lens OL, and is finally irradiated on the sample 7. The amount of the current of the electron beam 9 irradiating on the sample 7 changes depending on the rate of the electron beam 9 passing through the opening 2d of the beam control panel 2c to the entire electron beam 9. FIGS. 5A and 5B are plan views each showing an electron beam 9 emitted (irradiated) on the beam control panel 2c. FIG. 5A shows an emission range 51 of the electron beam 9 at the measurement. In a case where an electron beam 9 having the amount of current I is uniformly emitted on the circular emission range 51, the amount of the current of the electron beam 9 passing through the opening 2d is expressed by (r12/r22) I where the radius of the opening 2d is r1 and the radius of the emission range 51 is r2. FIG. 5B shows a case of having where an emission range 52 of an electron beam 9 which is smaller than the emission range 51 in FIG. 5A. The total amount of the current of the electron beam remains unchanged regardless of the size of the emission ranges 51 or 52 of the electron beam. Therefore, in the case of having the smaller emission range 52 of the electron beam 9 as shown in FIG. 5B, the amount of the current of the electron beam 9 passing through the opening 2d is expressed by (r12/r32)I. This value is larger than that of the case shown in FIG. 5A. Accordingly, if the emission range of the electron beam 9 emitted on the beam control panel 2c is regulated as shown in FIG. 5B to be smaller than the emission range 51 for the measurement, the amount of the current of the electron beam 9 emitted on the sample 7 can be increased compared to the amount thereof for the measurement. FIG. 6 is a schematic diagram showing a track of an electron beam 9 generated at the stabilizing processing. As shown in FIG. 6, the focal distance d21 of the first condenser lens CL1 is made longer than the focal distance d11 of the first condenser lens CL1 for the measurement in order to narrow the emission range of the electron beam 9 emitted on the beam control panel 2c. The first condenser lens CL1 is formed of a magnetic coil. The value of excitation current supplied to the magnetic coil is made smaller than the value of excitation current supplied thereto for the measurement, in order to make the focal distance d21 of the first condenser lens CL1 longer than the focal distance d11 set for the measurement. A large amount of the electron beam 9 having passed through the opening 2d further undergoes convergence by the second condenser lens CL2. At this time, a focal distance d22 of the second condenser lens CL2 is regulated so as to be equal to the focal distance d12 of the second condenser lens CL2 for the measurement. Here, the focal distance d22 is regulated to be equal to the focal distance d12 for the measurement in a way that the value of excitation current supplied to a magnetic coil constituting the second condenser lens CL2 is made larger than the value of excitation current supplied thereto for the measurement. Setting the focal distance d22 of the second condenser lens CL2 to be equal to the focal distance d12 for the measurement allows the electron beam 9 to be focused in the same visual field as that for the measurement by not changing the magnification of the objective lens OL, and allows the amount of the current of the electron beam 9 emitted onto the sample 7 to be increased compared to that for the measurement. Regulating the values of the excitation currents to be supplied to the lenses in this way makes it possible to increase or decrease, easily at high speed, the amount of the current of the electron beam to be emitted on the sample. In the present embodiment, conditions to stabilize the potential on a sample surface are: the excitation current to be supplied to each of the first and the second condenser lenses CL1 and CL2; and a time period during which a current larger than that for the measurement is emitted onto the sample 7. Hereinbelow, a description will be given of a method of determining the stabilizing conditions with reference to the flowchart shown in FIG. 7. Firstly, in Step S11, reference dimensions are measured by use of a calibration sample. The calibration sample is made of the same material as the sample 7 which is a measurement target. The calibration sample is formed of a substrate which is entirely conductive, and the dimensions of the calibration sample are known in advance or will be measured in the first step. This calibration sample is grounded to be in a discharging state where the potential on the surface of the calibration sample is 0 [V]. Then, the dimensions of the calibration sample are measured and thereby set to be reference dimensions. The measurement is made by scanning the calibration sample while emitting (irradiating) the electron beam onto the calibration sample, and then obtaining an image from which the dimensions are obtained. In next Step S12, the calibration sample is disconnected from the ground. The disconnection of the calibration sample from the ground causes the sample surface to be charged positively or negatively. In next Step S13, the currents to be supplied respectively to the magnetic coils of the first and the second condenser lenses CL1 and CL2 are respectively set to predetermined values. For example, the current to be supplied to the magnetic coil of the first condenser lens CL1 is set to a value which is smaller than the current value set for the measurement, whereas the current to be supplied to the magnetic coil of the second condenser lens CL2 is set to a value which is larger than the current value set for the measurement. Here, the focal distance d22 is set to be equal to the focal distance d12 set for the measurement to maintain the same magnification condition of the objective lens OL. These settings allow the emission of a large amount of electron beam 9, for example, 10 times or more compared to that for the measurement. In next Step S14, in order to stabilize the charge on the sample surface, predetermined amounts of currents are respectively supplied to the magnetic coils of the first and the second condenser lenses CL1 and CL2. Under this setting, the electron beam 9 is emitted for a predetermined time period. Since a large amount of electron beam 9 is emitted at this time, the time period required for the emission can be considerably reduced to 1/10 or less, for example, of a time period conventionally required. In next Step of S15, the conditions for the electron beam 9 are switched back to the ones for the measurement, and then an electron microscope image for the calibration sample is obtained to measure the calibration sample. Specifically, the currents to be respectively supplied to the magnetic coils of the first and the second condenser lenses CL1 and CL2 for the measurement of the calibration sample are not the currents regulated at Step S13 and determined at Step S17, but the currents set for the measurement. Namely, the currents values are switched back to those for the measurement to obtain the electron microscope image of the calibration sample and then to measure the calibration sample. In next Step S16, it is determined whether or not the measured values of the calibration sample are stabilized. The measurement of the calibration sample is performed in Step S15 after the electron beam is emitted on the sample for a predetermined time period from the time point t=0 in order to stabilize the charge on the sample surface in Step S14. Then, in a case, for example, where the difference between a value measured at a time point t1 and a value measured at a time point t2 is within a tolerable variation, it is determined that the measured values are stabilized. Here, the time point t1 and the time point t2 are time points during the emission time period of the electron beam 9. Further in Step S16, the measured values are compared with the reference values obtained in Step S11. A difference (denoted by D) between each measured value and the corresponding reference value is obtained, and then it is determined whether or not the difference D is smaller than a predetermined value, for example, 0.2 [nm]. When it is determined that the difference D is smaller than the predetermined value, the emission time period is considered to be appropriate. Therefore, the emission time period tx of the electron beam 9 is stored in a storage 45 (FIG. 1), and then the processing proceeds to Step S17. On the other hand, when it is determined that the difference D is larger than the predetermined value, the processing proceeds to Step S11 to measure reference values again by use of the calibration sample. Then, the processing proceeds to Step S13 to change the amounts of the currents to be supplied to the first and the second condenser lenses CL1 and CL2. Then, the processing continues further. Note that the processing may proceed directly to Step S13, if desired, when it is determined that the difference D is larger than the predetermined value. In next Step S17, the stabilizing conditions are stored in the storage 45, and the processing is terminated. Here, the stabilizing conditions include: the value, set in Step S13, of the excitation current supplied to the magnetic coil of each of the first and the second condenser lenses CL1 and CL2; and the emission time period tx of the electron beam 9 obtained in Step S16. Here, the stabilizing conditions vary depending on a material of a used sample. Therefore, the stabilizing conditions for various materials of samples may be obtained and stored in the storage 45. As described above, in the charged particle beam inspection apparatus according to the present embodiment, the excitation current supplied to the first condenser lens is set to be smaller than that supplied thereto for the measurement, in order to increase the focal distance of the first condenser lens. Consequently, the area of the beam control panel on which the electron beam is emitted is narrowed so that the amount of the electron beam passing through the opening of the beam control panel is relatively increased compared to that for the measurement. In this way, the amount of current of the electron beam emitted (irradiated) on the sample is increased. The emission, for a predetermined time period, of a large amount of electron beam thus regulated makes it possible to stabilize the potential on a sample surface in a short time. As a result, a track of the electron beam is made steady, and dimensions of a pattern on the sample can be measured with accuracy. Additionally, in the present embodiment, the amount of current to be supplied to each of the first and the second condenser lenses (one of the stabilizing conditions) is obtained in advance in order to stabilize the above-described potential. Here, the amount of current corresponds to a desired time period to be spent for stabilizing. The stabilizing conditions for each sample are stored in the storage 45 shown in FIG. 1. Since the charged particle beam inspection apparatus automatically performs the stabilizing processing with reference to the stabilizing conditions stored in the storage, a user can save time and labor considerably. Next, with reference to a flowchart shown in FIG. 8, a description will be given of a method of measuring the sample 7 by with use of the charged particle beam inspection apparatus 100 of the present embodiment by stabilizing the potential on the sample 7. Here, it is assumed that a measuring point and stabilizing conditions for each measurement target sample are stored in the storage 45 in advance. Firstly, in Step S20, a sample 7 is brought in the sample chamber 16. In next Step S21, the controlling unit 20 extracts a measurement point for the sample 7 from the storage 45, and moves an irradiation position to the measurement point by controlling the deflection coil 3. In next Steps S22 to S24, the surface of the sample 7 undergoes stabilizing processing. In Step S22, stabilizing conditions for the sample 7 are extracted from the storage 45. In Step S23, the value of the excitation current to be supplied to the first and the second condenser lenses CL1 and CL2 are regulated according to the values of the excitation currents included in the extracted stabilizing conditions. In next Step S24, an electron beam is emitted on the sample 7 according to an emission time period of the electron beam included in the extracted stabilizing conditions. In next Step S25 which is after the stabilizing processing, the value of the excitation current to be supplied to each of the first and the second condenser lenses CL1 and CL2 is changed to the one for the measurement while the potential on the surface of the sample 7 is stabilized. Then, the sample 7 is measured. Steps S21 to S25 are repeated until all the measurement points in the sample 7 are measured. In next Step S26, the sample 7 for which the measurements have been completed is taken out of the sample chamber 16. In next Step S27, it is determined whether or not the measurement has been completed for all samples 7. If all the measurements have not been completed, the processing returns to Step S20 to continue the remaining measurements. As described above, in the measurement method by use of the charged particle beam inspection apparatus of the present embodiment, the potential of the sample surface around the measurement point on the sample is stabilized prior to the measurement of the dimensions of the pattern formed on the sample. Before measuring the sample, the charged particle beam inspection apparatus automatically performs optimal stabilizing processing on the potential of each sample with reference to the stabilizing conditions stored in the storage 45 in advance. This processing stabilizes the potential of the sample surface in a short time as desired. Therefore, the electron beam is not affected by various different potentials on the sample, so that the electron beam does not change its emission range. Thus, this measurement method enables a steady measurement. Furthermore, a user can save time and labor considerably, because the charged particle beam inspection apparatus automatically performs the stabilizing processing with reference to the stabilizing conditions stored in the storage 45. Note that, in the present embodiment, the descriptions are given of a case where the charged particle beam inspection apparatus is applied to the measurement of dimensions of a pattern. However, the application of the charged particle beam inspection apparatus is not limited to this example. For example, the charged particle beam inspection apparatus may be applied to a defect detection of an element pattern formed on a semiconductor substrate, for example. In addition, in the present embodiment, the calibration sample is used to obtain the emission time period required for stabilizing the potential of the sample surface. However, the emission time period may be firstly obtained using the measurement target sample 7 in place of the calibration sample. In this case, the emission time period is determined in the following way. Specifically, the electron beam is emitted on the sample 7 to obtain the measurement values, and then a time period during which the relative measurement variation between each two of the multiple measurement values thus obtained is within a tolerable variation is set to be the emission time period. Further, the emission region to be stabilized may be made smaller than the measurement region of the measurement target sample 7, for example, a minimum size in such a way that the difference between the measured value and the reference value falls within a range of tolerable variation. With this emission region, the emission time period required for stabilizing can be minimized. Furthermore, during the stabilization of the potential on the surface of the sample 7, focusing conditions for the first and the second condenser lenses CL1 and CL2 and the objective lens CL3 may be controlled so that the electron beam can be focused on the sample 7. Alternatively, focusing conditions for the first and the second condenser lenses CL1 and CL2 and the objective lens CL3 may be controlled so that the electron beam can be diffused on the sample 7. In the latter case, the sample 7 is prevented from being locally damaged. |
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047284870 | description | DETAILED DESCRIPTION OF THE INVENTION In the following description, like reference characters designate like or corresponding parts throughout the several views of the drawings. Also, in the following description, it is to be understood that such terms as "forward", "rearward", "left", "right", "upwardly", "downwardly", and the like, are words of convenience and are not to be construed as limiting terms. IN GENERAL Referring now to the drawings, and particularly to FIG. 1, there is shown an elevational view of a fuel assembly, represented in vertically foreshortened form and being generally designated by the numeral 20. The fuel assembly 20 is the type used in a pressurized water reactor (PWR) and basically includes a lower end structure or bottom nozzle 22 for supporting the assembly on the lower core plate (not shown) in the core region of a reactor (not shown), and a number of longitudinally extending guide tubes or thimbles 24 which project upwardly from the bottom nozzle 22. The assembly 20 further includes a plurality of transverse grids 26 axially spaced along the guide thimbles 24 and an organized array of elongated fuel rods 28 transversely spaced and supported by the grids 26. Also, the assembly 20 has an instrumentation tube 30 located in the center thereof and an upper end structure or top nozzle 32 attached to the upper ends of the guide thimbles 24. With such an arrangement of parts, the fuel assembly 20 forms an integral unit capable of being conventionally handled without damaging the assembly parts. As mentioned above, the fuel rods 28 in the array thereof in the assembly 20 are held in spaced relationship with one another by the grids 26 spaced along the fuel assembly length. Each fuel rod 28 includes nuclear fuel pellets (not shown) and the opposite ends of the rod are closed by upper and lower end plugs 34, 36. The fuel pellets composed of fissile material are responsible for creating the reactive power of the PWR. A liquid moderator/ coolant such as water, or water containing boron, is pumped upwardly through the fuel assemblies of the core in order to extract some of the heat generated therein for the production of useful work. STANDARDIZED REDUCED LENGTH BURNABLE ABSORBER RODS In the operation of a PWR it is desirable to prolong the life of the reactor core as long as feasible to better utilize the uranium fuel and thereby reduce fuel costs. To attain this objective, it is common practice to provide an excess of reactivity initially in the reactor core and, at the same time, provide means to maintain the reactivity relatively constant over its lifetime. The present invention relates to such means in the form of a cluster of standardized reduced length burnable absorber rods 38, as seen in FIG. 1, inserted in the guide thimbles 24. The cluster of standardized rods 38 are stationarily supported by a holddown assembly 40 in the guide thimbles 24 of some of the fuel assemblies 20 to assist the movable control rods (not shown) in the guide thimbles 24 of other fuel assemblies (not shown) in maintaining a substantially constant level of neutron flux or reactivity in the core throughout its operating cycle. Heretofore, burnable absorber or poison rods have lately been of a custom-designed reduced length construction for maximization of the peaking factor margin in the reactor core. However, these latter reduced length rods have created other problems in their manufacture and handling, as discussed earlier. The standardized reduced length rod 38 of the present invention has substantially eliminated the drawbacks of the prior art rod while retaining its advantages. Referring now to FIGS. 2 and 3, in addition to FIG. 1, the holddown assembly 40 which supports the standardized rods 38 in spaced side-by-side relationship includes a lower flat perforated support plate 42 which fits within the fuel assembly top nozzle 32 and rests on the adapter plate 44 of the top nozzle. The holddown assembly 40 also includes a sleeve 46, being attached at its lower end within a central opening 48 in the support plate 42 and extending upwardly therefrom, and an upper holddown plate 50 which receives the sleeve 46 and is slidable vertically along it. Further, a holddown coil spring 52 is disposed about the sleeve 46 and extends between the lower support plate 42 and the upper holddown plate 50. Thus, the support plate 42 is held down against the top nozzle adapter plate 44 by the coil spring 52 which is compressed by the upper core plate (not shown) acting through the upper holddown plate 50 which abuts the upper core plate. This arrangement assures that the standardized rods 38 which are attached to the support plate 42 cannot be ejected from the reactor core by coolant flow forces while any thermal growth of the rods is accommodated. Turning next to FIG. 4, there is shown one embodiment of the standardized reduced length burnable absorber rod 38 of the present invention. The standardized rod 38 basically includes an upper end plug 54 having a threaded upper end 56 by which means the rod is connected to the support plate 42, a lower end plug 58, an elongated middle tubular rod section 60 having opposite upper and lower ends 62, 64 and a chamber 66 defined therein between the opposite ends 62, 64, an upper end spacer section 68 in the form of an elongated tubular extension extending between and rigidly interconnecting the upper end plug 54 and the upper end 62 of the middle tubular section 60, and a lower end spacer section 70 in the form of an elongated tubular extension extending between and rigidly interconnecting the lower end plug 58 and the lower end 64 of the middle tubular section 60. The rigid interconnection can take any suitable form, for instance, screw threaded, welded or crimped connections. By way of example, the sections 68, 60, 70 are shown welded together at W in FIGS. 4-7. Also, the standardized rod 38 has a burnable absorber material 72 disposed in the chamber 66 of the middle tubular section 60 and a pair of upper and lower end caps 74, 76 which seal the chamber 66. Specifically, the upper end cap 74 is spaced below the upper end plug 54 and attached in any suitable manner, such as by welding, to the upper end 62 of the middle tubular section 60 so as to seal the chamber 66 at the upper end 62, while the lower end cap 76 is spaced above the lower end plug 58 and attached also in any suitable manner, such as by welding, to the lower end 64 of the middle tubular section 60 so as to seal the chamber 66 at the lower end 64. Also, preferably a coil spring 78 is disposed in the chamber 66 of the middle tubular section 60 between the upper end cap 74 and the burnable absorber material 72 contained in the chamber 66 for maintaining the material in a stationary position within the chamber. In the one embodiment of the standardized rod 38 of FIGS. 4 and 8, the burnable absorber material 72 is in the form of a borosilicate glass tube 80 disposed in the chamber 66 of the middle tubular section 60 between its upper and lower end caps 74, 76. The upper end spacer section or tubular extension 68 of the standardized rod 38 has a hollow chamber 82 formed therein which defines an upper space extending on the upper end 62 of the middle tubular section 60 of the rod 38. Similarly, the lower end spacer section or tubular extension 70 of the standardized rod 38 has a hollow chamber 84 formed therein which defines a lower space extending between the lower end plug 58 and the lower end cap 76 of the middle tubular section 60 of the rod 38. The purpose of the spacer sections 68, 70 will become clear hereinafter. Another embodiment of the standardized rod 38' is illustrated in FIGS. 5 and 9. It only differs from the construction of the rod of FIGS. 4 and 8 in the particular burnable absorber material 86 being used therein, such being in the form of a stack of pellets 88 containing a boron compound, B.sub.4 C, and the specific axial lengths of the three sections of the rod. All of the remaining parts of the rod of FIG. 5 which are substantially the same as those of FIG. 4 are identified with the same reference numeral followed by a single prime mark. The embodiments of FIGS. 6 and 7 are identified similarly with double and triple prime marks being used in the respective figures. With respect to the standardized rods 38, 38', 38", 38"' of FIGS. 4 to 7, their upper and lower end spacer sections 68,70; 68',70'; 68",70"; and 68"',-70"' and the middle tubular sections 60,60',60",60"' each has one of a plurality of known different standard axial lengths being selected so that the upper and lower tubular extensions and middle tubular sections when rigidly interconnected together to form any one rod will have a combined axial length which is the same from rod to rod. However, in view that the axial lengths of the middle sections 60,60', 60",60"' of each rod can differ, the rod in effect simulates the prior art custom-designed reduced length rod. The upper and lower end spacer sections are selected to have certain axial lengths in view of the length of the middle section in order to ensure that the overall axial length of the rod will meet the desired standard length. Thus, the end spacer sections merely define empty space of varying axial lengths at the opposite ends depending on the length of the burnable absorber material in the middle section. In FIGS. 4 to 8, none of the axial lengths a, d, g, j of the upper end spacer sections 68,68',68",68"' of the standardized rods are equal. On the other hand, the axial lengths b and e of the middle sections 60, 60' are equal as are the axial lengths h and k of the middle sections 60", 60"'. Finally, the axial lengths c and i of the combined lower end spacer sections and lower end plugs 70,58 and 70",58" are equal as are the axial lengths f and i of the combined lower end spacer sections and lower end plugs 70',58' and 70"',58"'. With such variability in the axial lengths of the individual sections making up the rods 38 in a cluster thereof, it is readily seen that the axial location of the burnable absorber material can be different from rod to rod, although not necessarily different for every one of the rods. Each rod can be fitted with upper and rigidly lower end spacer sections of appropriate axial lengths selected from a number of different standard axial lengths available so that together with a middle section of any one of several different standard axial lengths, an overall rod is assembled having the same length as every other rod regardless of its absorber length or placement. With respect to adjacent rods, the upper and lower ends of the middle sections which contain the absorber material do not normally align with one another and thus provide a "ragged edge" at both the upper and lower ends. The "ragged edge" of the standardized rods of the present invention provides a smoother transition of material characteristics if the maximum peaking factor should occur at such interfaces than in the case of the aligned edge of the custom-fit reduced length rods of the prior art, which provides a significant reduction in the peaking power. FIGS. 10 and 11 show two different patterns of mixed clusters of standardized rods having burnable absorber middle sections of two different axial lengths. In FIG. 10, circles marked with an "x" signify rods having absorber material of one reduced length, such as 104 inches (or 264.16 centimeters), and circles marked with a dot "." signify rods having absorber material of another reduced length, such as 134 inches (or 340.36 centimeters), whereby the effective average axial length of the absorber material across the cluster is 108 inches (or 274.32 centimeters). In FIG. 11, circles marked with an "x" signify rods having absorber material, for instance, of an axial length of 108 inches (or 274.32 centimeters), whereas circles marked with a dot "." signify rods having absorber material of 120 inches (or 304.8 centimeters) in axial length, for an effective average axial length of 110 inches (or 279.4 centimeters) of absorber material across the cluster. It is thought that the standardized reduced length burnable absorber rod of the present invention and many of its attendant advantages will be understood from the foregoing description and it will be apparent that various changes may be made in the form, construction and arrangement thereof without departing from the spirit and scope of the invention or sacrificing all of its material advantages, the form hereinbefore described being merely a preferred or exemplary embodiment thereof. |
claims | 1. An incore piping section maintenance system of a reactor, comprising: a maintenance system main body having a portion to be inserted into a pipe of an incore piping section located in a reactor pressure vessel; a main body supporting mechanism for detachably fixing the maintenance system main body to an inner surface of the pipe; a turning arm mounted to another portion of the maintenance system main body and disposed outside the pipe and the maintenance system main body; a turning mechanism for turning and driving the turning arm; an axial moving mechanism disposed to the turning arm and slidable in a direction substantially parallel to a central axis of the pipe; a laser generator for generating a laser beam; a laser de-sensitization treatment mechanism supported by the axial moving mechanism and for irradiating the laser beam to an outer surface of the pipe; and an optical transmission mechanism for guiding the laser beam outputted from the laser generator to the laser de-sensitization mechanism. 2. The incore piping section maintenance system of a reactor according to claim 1 , wherein the incore piping section maintenance system includes at least three main body supporting mechanisms and each of said main body supporting mechanisms is constructed in combination with a link mechanism including a guide member and a cylinder assembly for driving the guide member of the link mechanism in and out of the maintenance system main body. claim 1 |
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claims | 1. An apparatus for managing test procedures for a hardware-in-the-loop (HIL) simulation environment, the apparatus comprising:an input interface for receiving input from a user; anda first processor coupled to the input interface and in operable communication with the HIL simulation environment, the first processor configured to:generate a test sequence comprising a plurality of test procedure references based on input from the user, wherein each test procedure reference corresponds to a test procedure that comprises instructions for issuing commands to, and receiving data from, the HIL simulation environment, and wherein each test sequence comprises instructions for creating a desired operating condition for the simulation environment using operational data and diagnostic data; andsequentially execute each referenced test procedure within the generated test sequence in cooperation with the HIL simulation environment, in response to a command from the user. 2. The apparatus of claim 1, wherein the first processor is further configured to:insert at least one test command within the generated test sequence based on input from the user, wherein each test command is associated with a predetermined operation; andsequentially execute each test command and each referenced test procedure within the test sequence in cooperation with the HIL simulation environment, in response to the command from the user. 3. The apparatus of claim 2, wherein the processor is further configured to store the test sequence to be retrieved at a later time. 4. The apparatus of claim 2, wherein:the at least one test command comprises a wait command that is associated with a wait period; andthe first processor is further configured to pause for a period of time that is substantially equally to the wait period before executing the next test command or referenced test procedure in the test sequence in response to the wait command. 5. The apparatus of claim 2, further comprising a display device that is coupled to the processor and wherein:the at least one test command comprises a prompt command that is associated with a message and a timeout period; andthe first processor is further configured to display the message to the user and to pause during a period of time that is substantially equal to the timeout period before proceeding to the next test command or referenced test procedure in the test sequence in response to the prompt command. 6. The apparatus of claim 2, wherein:the HIL simulation environment includes an HIL simulator having a second processor configured to load and execute a plant model;the at least one test command further comprises a reload plant model command; andthe first processor is further configured to cause the second processor to reload the plant model in response to the reload plant model command. 7. The apparatus of claim 2, wherein:the HIL simulation environment includes an HIL simulator having a second processor configured to execute a plant model that includes a plurality of environment variables;the at least one test command further comprises a clear environment variables command; andthe first processor is further configured to cause the second processor to assign default values to the plurality of environment variables in response to the clear environment variables command. 8. The apparatus of claim 2, wherein:the HIL simulation environment includes an HIL simulator;the at least one test command further comprises a reboot HIL simulator command; andthe first processor is further configured to cause the HIL simulator to reboot in response to the reboot HIL simulator command. 9. The apparatus of claim 2, wherein:each referenced test procedure comprises instructions written in a scripting language; andthe first processor is further configured to access an interpreter software module to execute the instructions. 10. The apparatus of claim 2, wherein each referenced test procedure further comprises instructions for generating a test procedure log that includes describes the commands issued to, and the data received from, the HIL simulation environment during execution of the referenced test procedure. 11. The apparatus of claim 10, wherein the processor is further configured to compile a summary test sequence report describing the execution of each reference test procedure within the test sequence based on the test procedure log for each referenced test procedure. 12. The apparatus of claim 2, further comprising a display device coupled to the first processor and configured to display a user interface comprising:a test sequence display area for displaying the at least one test command and the plurality of test procedure references within the test sequence;a selectable add test procedure reference control for adding a test procedure reference to the test sequence;a test command control for adding a test command to the test sequence; anda selectable test sequence performance control for causing the first processor to sequentially execute the at least one test command and the plurality of referenced test procedures. 13. A non-transitory computer readable medium comprising instructions that, when executed by a processor that is in operable communication with a hardware-in-the-loop (HIL) simulation environment, cause the processor to perform the following steps:generating a test sequence comprising a plurality of test procedure references and a plurality of test commands, wherein:each test procedure reference corresponds to a test procedure comprising instructions for performing at least one test within the HIL simulation environment by issuing commands to, and receiving data from, the HIL simulation environment;each test command is associated with a predetermined operation; andeach test sequence comprises instructions for creating a desired operating condition for the simulation environment using operational data and diagnostic data; andsequentially executing each referenced test procedure and each command within the test sequence in cooperation with the HIL simulation environment. 14. The computer readable medium of claim 13, wherein each referenced test procedure further comprises instructions written in a scripting language and the step of sequentially executing includes accessing an interpreter software module to execute the instructions. 15. The computer readable medium of claim 13, wherein each referenced test procedure is further comprises instructions for generating a test procedure log comprising data describing the commands issued to the HIL simulation environment, the data received from the HIL simulation environment, and the results of the at least one test performed during the execution of the referenced test procedure. 16. The computer readable medium of claim 15, wherein the method further comprises generating a test sequence summary report describing the execution of each referenced test procedure within the test sequence based on the test procedure log for each referenced test procedure. 17. A system for automating the performance of a plurality of test procedures, the system comprising:a hardware-in-the-loop (HIL) simulation environment that comprises an HIL simulator and an electronic system under test; anda host electronic device in operable communication with the HIL simulation environment and comprising:an input interface for receiving input from a user; anda processor coupled to the input interface and configured to:generate a test sequence comprising a plurality of test procedure references and test commands based on input from the user, wherein each test procedure reference corresponds to a test procedure comprising instructions for issuing commands to, and receiving data from, the HIL simulation environment and each test command is associated with a predetermined operation, and wherein each test sequence comprises instructions for creating a desired operating condition for the simulation environment using operational data and diagnostic data; andsequentially execute each referenced test procedure and each test command within the test sequence in cooperation with the HIL simulation environment, upon receiving a command from the user. 18. The system of claim 17, wherein each test procedure further comprises instructions for generating a test procedure log that describes the commands issued to, and the data received from, the HIL simulation environment during execution of the referenced test procedure. |
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059498390 | summary | TECHNICAL FIELD The present invention relates to a fuel assembly for a light-water nuclear reactor, and more specifically of boiling-water type. BACKGROUND OF THE INVENTION A core in a boiling water reactor comprises a plurality of vertically arranged fuel assemblies. A fuel assembly comprises a bundle of vertical elongated fuel rods which are retained and fixed by a number of spacers arranged in spaced relationship to each other along the bundle. The spacers comprise a number of cells for mutually fixing the fuel rods. The ends of the bundle are retained at the bottom and at the top, by a bottom tie plate and a top tie plate. In certain fuel assemblies, the bundle is divided into four orthogonal sub-bundles and the lower and upper parts, of each sub-bundle are retained by a bottom tie plate and a top tie plate, respectively. The bundle and the sub-bundles, are surrounded by a fuel channel which is normally designed with a square cross section. The fuel assembly comprises a vertical channel containing non-boiling water surrounded by a tubular casing. In the following, the channel including its casing will be referred to as a water channel. The water channel extends through the whole fuel assembly and has a cross section which may be circular or cruciform. The fuel rods contain a stack of pellets of a nuclear fuel arranged in a cladding tube. During the burnup of the nuclear fuel, fission gases contained inside the fuel rod are released. To prevent the pressure on the cladding from becoming too great, an expansion space for the fission gases is needed, a so-called fission gas plenum. The fission gas plenum is normally arranged above the stack of fuel pellets. The core is immersed into water which serves both as a coolant and as a neutron moderator. During operation, the water flows from below and upwardly through the fuel assembly, whereby part of the water is transformed into steam. The proportion of steam is highest in the upper part of the fuel assembly. In the following, coolant means the water and the steam which flow through the fuel assembly. When the coolant flows upwardly through the fuel assembly, it is important that it be subjected to as little pressure drop as possible. Because of the high proportion of steam, the pressure drop is higher in the upper part of the fuel assembly than in the lower part thereof. During operation, the pressure drop across a spacer in the upper part of the fuel assembly is about five times higher than the pressure drop across a corresponding spacer in the lower part of the fuel assembly. Therefore, it is of particular importance to design the upper part of the fuel assembly in such a way that there is as low a pressure drop as possible. A low pressure drop in the upper part of the fuel assembly is favorable for the stability properties of the fuel assembly and reduces the risk of dryout. A high pressure drop means a high pressure on the fuel channel and may give rise to creeping and subsequently unacceptable deformation of the fuel channel. SUMMARY OF THE INVENTION The object of the invention is to provide a fuel assembly which, during operation, has a low pressure drop in its upper part. A fuel assembly for a boiling water reactor, wherein the fuel assembly during operation is arranged vertically in the core of the reactor. The fuel assembly comprises a plurality of vertical fuel rods arranged in at least one fuel bundle, wherein each one of the fuel rods comprises a stack of fuel pellets surrounded by a cladding tube. A bottom tie plate retains and supports the lower part of the fuel bundle and is arranged at the bottom of the fuel assembly. A number of spacers retain and position the fuel rods in spaced relationship to each other and are arranged axially separated along the fuel rods. A vertical water channel is provided through which water flows upwards through the fuel assembly, and a fuel channel which surrounds the fuel bundle. A top spacer which retains and supports the upper part of the fuel bundle and is arranged above and spaced apart from the stacks of fuel pellets, wherein the majority of the fuel rods are provided with a plenum tube which is connected to the upper part of the cladding tube and which surrounds a fission gas plenum, the plenum tube having a cross-section area which is smaller than the cross-section of the cladding tube. |
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044143397 | abstract | Low density compositions containing colloidal-size particulates of an ELM absorber such as Fe.sub.3 O.sub.4 and a particulate of an ELM attenuator such as carbonyl iron dispersed in a dielectric matrix such as a styrene/butyl acrylate copolymer provide improved absorption of electromagnetic radiations. |
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