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description | This application claims the benefit of Korean Patent Application No. 10-2010-0085730, filed on Sep. 1, 2010, in the Korean Intellectual Property Office, the disclosure of which is incorporated herein by reference. 1. Field of the Invention The present invention relates to a prefabricated vol-oxidizer for spent nuclear fuel, enabling convenient operation and maintenance thereof. 2. Description of the Related Art Nuclear fuel is a substance capable of extracting a usable energy by continuously causing nuclear fission by being fed in a nuclear reactor. Spent nuclear fuel refers to a used substance remaining after the nuclear fission. Such spent nuclear fuel may be managed generally by two methods. One method is to thoroughly isolate the spent nuclear fuel from the human ecosystem by burying the spent nuclear fuel in a rock bed lying at about 500 m or more underground, which is called permanent disposal. The other method is to reuse the nuclear fuel by extracting a recyclable substance from the spent nuclear fuel while permanently disposing of high-level radioactive substances. According to those conventional methods, a spent nuclear fuel assembly combusted in a nuclear power plant is no longer processed and stored in a water reservoir. However, as a driving time of the nuclear power plant increases, quantity of used nuclear fuel rods gradually increases, consequently requiring a great storage space. Furthermore, risk and necessity for proper disposal of the accumulated nuclear waste are continuously raised as issues. Accordingly, there is a desire for a new scheme and method for managing and recycling spent nuclear fuel in a solid state. In this regard, an apparatus for partial processing that pulverizes and oxidizes the spent nuclear fuel for a following process has been developed. An aspect of the present invention provides a vol-oxidizer for spent nuclear fuel, capable of pulverizing rod-cuts using air and heat and enabling convenient maintenance and repair by a modularized structure. Another aspect of the present invention provides a vol-oxidizer for spent nuclear fuel, enabling convenient operation, maintenance, and repair in a remote manner from a highly-radioactive hot cell. According to an aspect of the present invention, there is provided a vol-oxidizer for spent nuclear fuel including a reactor module in which the spent nuclear fuel is placed and oxidized, a heating module to heat the reactor module to high temperature, a utility module to control an inner state of the reactor module, being connected with the reactor module, a support module to support the heating module and the reactor module, a drive module to supply a driving force for transferring the spent nuclear fuel within the reactor module, a valve module to control discharge of the spent nuclear fuel being oxidized in the reactor module, and a collecting container module to collect the spent nuclear fuel being oxidized and discharged from the reactor module, wherein the respective modules may be assembled and disassembled with respect to one another. According to embodiments of the present invention, separation and connection of a vol-oxidizer for spent nuclear fuel may be conveniently performed in a remote manner. Additionally, according to embodiments of the present invention, maintenance and repair of the vol-oxidizer for spent nuclear fuel are facilitated. Reference will now be made in detail to exemplary embodiments of the present invention, examples of which are illustrated in the accompanying drawings, wherein like reference numerals refer to the like elements throughout. Therefore, the description about a certain drawing may be made with reference to the other drawings. Features generally known in the art or repeatedly explained throughout will be omitted. FIG. 1 is a perspective view illustrating a vol-oxidizer 1 for spent nuclear fuel according to an embodiment of the present invention. FIG. 2 is another perspective view of the spent nuclear fuel vol-oxidizer 1 of FIG. 1. Referring to FIGS. 1 and 2, the spent nuclear fuel vol-oxidizer 1 includes a heating module 10, a reactor module 20, a drive module 30, a utility module 40, a support module 50, an air cylinder module 60, a valve module 70, a collecting container module 80, a terminal block module 90, and a guide module 95. The spent nuclear fuel vol-oxidizer 1 may oxidize spent nuclear fuel by heating the spent nuclear fuel to a high temperature under vacuum. The heating module 10 supplies heat for oxidizing the spent nuclear fuel. The spent nuclear fuel may be put in the reactor module 20, where the reactor module 20 then oxidizes the spent nuclear fuel. The drive module 30 may supply a driving force for transferring the spent nuclear fuel within the reactor module 20. The support module 50 may support the heating module 10, the reactor module 20, and the drive module 30. The valve module 70 controls discharge of the spent nuclear fuel oxidized by the reactor module 20. The collecting container module 80 may collect the spent nuclear fuel being oxidized and discharged from the reactor module 20. The guide module 95 may guide movement of the heating module 10. The spent nuclear fuel vol-oxidizer 1 is prefabricated. That is, the spent nuclear fuel vol-oxidizer 1 may be connected and separated in units of the respective modules. Possibility of remote connection and separation of the modules may be analyzed in view of visibility, interference, approach, weight, and the like. The heating module 10, the utility module 40, the drive module 30, the valve module 70, and the collecting container module 80 may be made of SUS304 and ceramic. FIG. 3A is a perspective view of the heating module 10 of the spent nuclear fuel vol-oxidizer 1. FIG. 3B is a front view of the heating module 10. Referring to FIGS. 1 through 3B, the heating module 10 may heat the reactor module 20 to a high temperature so as to oxidize the spent nuclear fuel. More specifically, the heating module 10 may heat an oxidation container provided in the reactor module 20 up to approximately 500 Celsius degrees or more to oxidize the spent nuclear fuel. For this purpose, according to an exemplary embodiment, the heating module 10 may be made of a material having a low thermal expansion coefficient so as not to change much in shape, even at the high temperature. The heating module 10 may include a first heating body 11 and a second heating body 12. The first heating body 11 may form a left half portion of the heating module 10 while the second heating body 12 forms a right half portion. The first heating body 11 and the second heating body 12 may be symmetrically disposed on the left and the right with respect to a vertical bisector of the heating module 10. The first heating body 11 and the second heating body 12 may be connected at both sides with respect to the reactor module 20 disposed therebetween. A space portion for receiving the reactor module 20 is depressed inside the first heating body 11 and the second heating body 12. According to this structure, the heating module 10 surrounds the reactor module 20. Additionally, a heating member H such as an electric heating wire may be provided on inner surfaces of the first heating body 11 and the second heating body 12. The heating module 10 may further include an introduction unit hole 121 fit around an introduction unit 22 of the reactor module 20, and a rotation shaft hole 14 fit around a rotation shaft 35 of the drive module 30. The first heating body 11 and the second heating body 12 may be mounted on the support module 50 in a sliding manner. The first heating body 11 and the second heating body 12 may linearly slide on the support module 50. For example, the first heating body 11 and the second heating body 12 may slide in lateral directions. Pulling tension may be applied to the first heating body 11 and the second heating body 12 in the lateral directions, for separation of the heating module 10. Movements of the first heating body 11 and the second heating body 12 may be guided by the guide module 95. A handling tool connector 13 may be provided to an upper end of the heating module 10 for connection with a handling tool such as a hook or a crane. The handling tool connector 13 may facilitate remote connection and separation of the heating module 10. The handling tool connector 13 may include a hole to efficiently connect the handling tool. Hereinafter, processes of remotely connecting and separating the heating module 10 will be described. First, to separate the heating module 10 from the support module 50, the first heating body 11 is pulled to the left and the second heating body 12 is pulled to the right so that the first heating body 11 and the second heating body 12 are separated from each other. The first heating body 11 and the second heating body 12 are slid on the support module 50 in the opposite directions and separated by a predetermined interval from each other. Next, the handling tool such as a hook or a crane is connected to the handling tool connector 13 disposed at upper ends of the first heating body 11 and the second heating body 12. The handling tool lifts the first heating body 11 and the second heating body 12 from the support module 50. On the other hand, to remotely connect the heating module 10 to the support module 50, the handling tool such as a hook or a crane is connected to the handling tool connector 13 of the first heating body 11 and the second heating body 12. The first heating body 11 and the second heating body 12 are mounted on the support module 50. More specifically, the first heating body 11 and the second heating body 12 may be mounted on the guide module 95. After the first heating body 11 and the second heating body 12 are mounted on the guide module 95, a pushing force is applied to the first heating body 11 and the second heating body 12 in a direction toward the reactor module 20 disposed between the first heating body 11 and the second heating body 12, thereby connecting the first heating body 11 and the second heating body 12 to each other. Since the first heating body 11 and the second heating body 12 are connected with the reactor module 20 disposed therebetween, the support module 50 is surrounded by the first heating body 11 and the second heating body 12. FIG. 4 is a perspective view of the reactor module 20 of the spent nuclear fuel vol-oxidizer 1 of FIG. 1. Referring to FIGS. 1, 2, and 4, the reactor module 20 may include a body unit 21 and the introduction unit 22. The oxidation container adapted to receive the spent nuclear fuel may be mounted in the body unit 21. The oxidation container may have a hollow cylindrical shape. A circumferential surface of the oxidation container may be formed of mesh. Due to the mesh structure, the spent nuclear fuel before being oxidized may be disposed in the oxidation container. However, when the spent nuclear fuel is separated into hulls and oxidation powders, the hulls remain in the oxidation container whereas the oxidation powders are passed through the mesh and collected by a powder receiver (not shown) disposed at a lower portion of the oxidation container. The powder receiver may be rotatable by about 180 degrees from a lower side to an upper side of the oxidation container with respect to the rotation shaft 35. Accordingly, the oxidation powder collected in the powder receiver may be transferred to a powder transfer unit 212 that will be described hereinafter. Additionally, the body unit 21 includes a powder transfer unit 212 and the hull transfer unit 214 disposed at a lower portion of the body unit 21. The hull transfer unit 214 may transfer the hulls separated in the oxidation container to the collecting container module 80. Specifically, the hulls left in the oxidation container by the oxidation are transferred to the collecting container module 80 by the hull transfer unit 214. The powder transfer unit 212 may transfer the oxidation powder separated from the oxidation container. To be more specific, when the oxidation powder are collected to the powder receiver after the oxidation, the powder receiver is rotated, thereby transferring the oxidation powder to the powder transfer unit 212. The oxidation powder transferred to the powder transfer unit 212 are then transferred to the collecting container module 80. The powder transfer unit 212 may be sloped with respect to a horizontal plane by a predetermined angle, for example, by about 45 degrees. The introduction unit 22 having a tubular shape may introduce the spent nuclear fuel into the oxidation container from the outside. One end of the introduction unit 22 protrudes out of the oxidation container. Specifically, the introduction unit 22 may have a hollow cylindrical shape and supply a path for movement of the spent nuclear fuel from an outside of the body unit 21 up to the oxidation container. The body unit 21 may be connected with a decompression unit 41, a degassing and decompression tube 42, an oxidizing agent supply unit (not shown), a pressure gauge 44, and a temperature sensor connector 43. A pressure gauge fastening clamp 441, a temperature sensor fastening clamp 421, and a decompression tube fastening clamp 411 may be further connected, which will be described in detail hereinafter. FIG. 5 is a perspective view of the utility module 40 of the spent nuclear fuel vol-oxidizer 1 of FIG. 1. Referring to FIGS. 1, 2, and 5, the utility module 40 may be connected to the reactor module 20 to control conditions of the reactor module 20. The utility module 40 may include the decompression unit 41, the degassing and decompression tube 42, the oxidizing agent supply unit (not shown), and an introduction unit cover 45. The decompression unit 41 and the degassing and decompression tube 42 may enable the reactor module 20 to perform the high temperature oxidation in a decompressed state. More specifically, the decompression unit 41 and the degassing and decompression tube 42 may be connected to a decompression pump (not shown) installed at an outside of the reactor module 20. The decompression pump may decompress the reactor module 20 to a vacuum state so that the high temperature oxidation is performed under vacuum. One end of the decompression unit 41 is connected to an inside of the body unit 21 while the other end of the decompression unit 41 is protruded out of the body unit 21. The decompression unit 41 is connected with the body unit 21 and may pass through the body unit 21. The decompression unit 41 may include a decompression valve (not shown) adapted to control an inner pressure of the body unit 21. One end of the degassing and decompression tube 42 is connected to the decompression unit 41 and the other end is connected to the decompression pump (not shown) installed at an outside of the body unit 21. The degassing and decompression tube 42 may be bent at a position between the decompression unit 41 and the decompression pump. The degassing and decompression tube 42 is connected to the body unit 21 and may pass through the body unit 21. Therefore, volatile gas such as Kr, I, tritium (H3), and the like generated from the spent nuclear fuel during the high-temperature oxidation may be removed by the degassing and decompression tube 42. One end of the temperature sensor connector 43 may be connected to the inside of the body unit 21 whereas the other end is exposed to the outside of the body unit 21. The degassing and decompression tube 42 may include a valve (not shown) adapted to selectively remove gases generated according to temperatures during the high-temperature oxidation. The oxidizing agent supply unit (not shown) may supply a path for supply of an oxidizing agent into the oxidation container. The pressure gauge 44 may be connected to the body unit 21 by passing through the body unit 21. Generally, oxygen (O2) is used as the oxidizing agent. The oxidizing agent supply unit (not shown) may be in a tube shape having a predetermined diameter so as to be able to supply O2. The oxidizing agent supply unit (not shown) may include a valve to decompress an inside of the oxidation container to a vacuum state during the high-temperature heating. The heating module 10 may include holes for the decompression unit 41, the degassing and decompression tube 42, and the pressure gauge 44 to pass through. The first heating body 11 and the second heating body 12 may include grooves having a curved cross section to form the holes. Therefore, the first heating body 11 and the second heating body 12 are connected to each other with the decompression unit 41, the degassing and decompression tube 42, and the pressure gauge 44 disposed therebetween. More specifically, the decompression unit 41, the degassing and decompression tube 42, and the pressure gauge 44 are mounted to the body unit 21 of the reactor module 20. Next, the first heating body 11 and the second heating body 12 are connected to each other with the degassing and decompression tube 42, the pressure gauge 44, and the reactor module 20 disposed therebetween. The introduction unit cover 45 is connected to one end of the introduction unit 22 of the reactor module 20. The introduction unit cover 45 may selectively open and close one end of the introduction unit 22, thereby controlling entry of the spent nuclear fuel. FIG. 6 is a perspective view of the drive module 30 of the spent nuclear fuel spent nuclear fuel 1 of FIG. 1. FIG. 7 is a plan view of the drive module 30 of the spent nuclear fuel vol-oxidizer 1. FIG. 8 is a front view of the drive module 30 of the spent nuclear fuel vol-oxidizer 1. Referring to FIGS. 1, 2, and 6 to 8, the drive module 30 may include a motor 31, a motor housing 32, power transmitters 33 and 34, the rotation shaft 35, and a transfer unit (not shown). A driving shaft of the motor 31 may be coplanar with the rotation shaft 35 and also be arranged perpendicular to the rotation shaft 35. A hook portion 311 may be provided at one side of the motor 31 to transfer the motor 31. When the handling tool such as a crane and a hook is fixed to the hook portion 311, the motor 31 may be lifted or transferred remotely. The motor 31 is separably connected to the motor housing 32. The motor 31 includes a wheel unit 312 for sliding connection with the motor housing 32. The wheel unit 312 may be rotatably connected to a lower end of the motor 31. The motor 31 may further include a stopper 313 to prevent the motor 31 from sliding by a predetermined distance away from the motor housing 32. The stopper 313 may be connected with a side surface or an upper surface of the motor 31 and protrude by a predetermined thickness from the motor 31. When the motor 31 is moved by more than the predetermined distance with respect to the motor housing 32, the stopper 313 is interfered with a part of the motor housing 32. Accordingly, the motor 31 is prevented from further moving with respect to the motor housing 32 and being separated from the motor housing 32. The motor housing 32 includes a space so that the motor 31 is received in the space. Additionally, the motor housing 32 includes a rail unit 322 for contact with the wheel unit 312. The rail unit 322 may be recessed to a predetermined depth to receive the wheel unit 312. The rail unit 322 may extend in a direction perpendicular to the rotation shaft 35. Accordingly, the driving shaft of the motor 31 may be perpendicular to the rotation shaft 35. The motor housing 32 is connected to the support module 50, for example by a fastening member. The motor housing 32 may be connected to the support module 50 to slope down toward the rotation shaft 35. Accordingly, the rail unit 322 of the motor housing 32 is sloped down toward the rotation shaft 35. Therefore, when the motor 31 is received in the motor housing 32, that is, when the wheel unit 312 is received in the rail unit 322, the motor 31 is slid toward the rotation shaft 35. The motor 31 is slid within the motor housing 32 until the stopper 313 is interfered with a part of the motor housing 32. Exemplarily, when sliding of the motor 31 is stopped by the stopper 313, the power transmitter 33 in connection with the driving shaft of the motor 31 is then connected to the power transmitter 34 in connection with the rotation shaft 35. Therefore, a connection portion between the motor 31 and the motor housing 32 may be configured to enable sliding of the motor 31, considering the connection structure between the power transmitters 33 and 34 and the slope structure at a surface receiving the motor 31. For example, the motor housing 32 may be sloped by about 5 degrees with respect to a horizontal plane. Thus, since connection and separation of the motor 31 may be remotely performed using the handling tool such as a crane, maintenance and repair of the motor 31 are facilitated. The driving force of the motor 31 may be transmitted to the rotation shaft 35 through the power transmitters 33 and 34. For example, bevel gears may be used as the power transmitters 33 and 34. In this case, the bevel gears may be disposed at an end of the driving shaft and an end of the rotation shaft 35. The rotation shaft 35 is disposed to pass through the oxidation container, the body unit 21, and the heating module 10. The one end of the rotation shaft 35 may be transmitted with the driving force from the motor 31 through the power transmitters 33 and 34. The transfer unit may be mounted to the rotation shaft 35. The transfer unit may be disposed in the oxidation container to transfer, into the oxidation container, the spent nuclear fuel supplied from the introduction unit 22. The transfer unit may have a screw shape and may be mounted to the rotation shaft 46 to be rotated integrally with the rotation shaft 35. FIGS. 9A and 9B are a perspective view and a side view, respective, of the support module 50 of the spent nuclear fuel vol-oxidizer 1 of FIG. 1. Referring to FIGS. 9A and 9B, the support module 50 may include a support board 51, a base unit 52, a support shaft 53, and a supporter unit 54. The support board 51 supports the heating module 10, the reactor module 20, and the drive module 30. The support board 51 may include a powder transfer unit hole 511 fit around the powder transfer unit 212, and a hull transfer unit hole 512 fit around the hull transfer unit 214. A powder transfer unit supporter 513 may be provided at one side of the powder transfer unit hole 511 to support the powder transfer unit 212. The powder transfer unit supporter 513 may be sloped down by a predetermined angle from the support board 51. For example, the powder transfer unit supporter 513 may be sloped corresponding to a slope angle of the powder transfer unit 212. The powder transfer unit supporter 513 may be formed by bending a part of the support board 51. Also, the powder transfer unit supporter 513 may have a thin plate shape. The support board 51 may include supporters 541 and 542 rotatably supporting the rotation shaft 35. The supporters 541 and 542 may be protruded from the support board 51 by a predetermined length. The supporters 541 and 542 may include a front supporter 541 to support a front end of the rotation shaft 35 and a rear supporter 542 to support a rear end of the rotation shaft 35. Bearings may be applied where the rotation shaft 35 is rotatably mounted to the supporters 541 and 542, for smoother rotation of the rotation shaft 35. The support shaft 53 is disposed at a lower part of the support board 51 to support the support board 51. The support shaft 53 may be extended downward from the support board 51 and rotatably connected to the base unit 52. As a result, the heating module 10, the reactor module 20, and the drive module 30 may be conveniently maintained and repaired. The base unit 52 forms a bottom surface of the support module 50. For this, the base unit 52 may have a wide plate shape capable of stably supporting the support module 50. The base unit 52 may include a support shaft connector 521 rotatably connected to the support shaft 53. The support shaft connector 521 may extend upward from the base unit 52. A bearing may be provided to the support shaft connector 521 so that the support shaft 53 is rotatably connected. More specifically, the support shaft connector 521 may have a larger diameter than the support shaft 53 so that the support shaft 53 is inserted in the support shaft connector 521. FIG. 10 is a perspective view of the air cylinder module 60 of the spent nuclear fuel vol-oxidizer 1. Referring to FIGS. 1, 2, and 10, the air cylinder module 60 may be connected to one side of the heating module 10 to supply a force for separate and connect the heating module 10. A first connector 61 for connection with the first heating body 11 may be provided to one end of the air cylinder module 60. A second connector 62 for connection with the second heating body 12 may be provided to the other end of the air cylinder module 60. The first connector 61 and the second connector 62 may be shaped as protrusions to be inserted in the first heating body 11 and the second heating body 12, respectively. The air cylinder module 60 may separate the heating module 10 by pushing the first heating body 11 and the second body 12 in opposite outward directions using a pneumatic pressure. Also, the air cylinder module 60 may connect the heating module 10 by pulling the first heating body 11 and the second body 12 toward each other. FIG. 11 is a front view of the valve module 70 of the spent nuclear fuel vol-oxidizer 1. Referring to FIGS. 1, 2, and 10, the valve module 70 may include a oxidation powder valve 71 to control discharge of the oxidation powder being transferred from the powder transfer unit 212, and a hull valve 72 to control discharge of the hulls being transferred from the hull transfer unit 214. Knife gate valves may be used as the oxidation powder valve 71 and the hull valve 72. In general, the knife gate valve refers to a device opening and closing flow of fluid, being connected between a tube member and another tube member. In particular, when the fluid has a predetermined viscosity and accordingly has friction and stagnation, the knife gate valve may be effective to control the flow of fluid. The oxidation powder valve 71 and the hull valve 72 may be connected and separated using clamps. FIG. 12 is a front view of the collecting container module of the spent nuclear fuel vol-oxidizer 1. Referring to FIGS. 1, 2, and 12, the collecting container module 80 may include a hull collecting container 81 and a oxidation powder collecting container 82 to collect the oxidation powder. The hull collecting container 81 and the oxidation powder collecting container 82 may be made of a flexible material or formed by welding. In addition, an end cap, which is slidable back and forth, may be separably connected to the hull collecting container 81 and the oxidation powder collecting container 82. FIG. 13 is a perspective view of the terminal block module 90 of the spent nuclear fuel vol-oxidizer 1. Referring to FIGS. 1, 2, and 13, the terminal block module 90 may be mounted to the support board 51 of the support module 50. The terminal block module 90 is connected to an external power supply and to an electric wire of the spent nuclear fuel vol-oxidizer 1. The terminal block module 90 may be connected to the support board 51 by a fastening member. The terminal block module 90 may include a plurality of holes. FIG. 14 is a perspective view of the guide module 95 of the spent nuclear fuel vol-oxidizer 1. Referring to FIGS. 1, 2, and 14, the guide module 95 may include a fixed unit 951 fixed to the support board 51, and a moving unit 952 connected to the fixed unit 951 to be movable linearly. The fixed unit 951 may extend in a predetermined direction and may include a guide portion to guide movement of the moving unit 952. For example, the guide portion may be a groove formed on the fixed unit 951. The fixed unit 951 may be fixed to the support module 50 by the fastening member. The moving unit 952 may be connected to bottom surfaces of the first heating body 11 and the second heating body 12 and moved integrally with the first heating body 11 and the second heating body 12. The moving unit 952 may be fixed to the first heating body 11 and the second heating body 12 by a fastening member. The movement of the moving unit 952 may be guided by the guide portion of the fixed unit 951. Although a few exemplary embodiments of the present invention have been shown and described, the present invention is not limited to the described exemplary embodiments. Instead, it would be appreciated by those skilled in the art that changes may be made to these exemplary embodiments without departing from the principles and spirit of the invention, the scope of which is defined by the claims and their equivalents. |
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abstract | The present invention provides an apparatus for detecting and/or repositioning annulus spacers used to maintain the position of a pressure tube within a calandria tube of a nuclear reactor. The method comprises the steps of: vibrationally isolating a section of the pressure tube; vibrating the wall of said pressure tube within said isolated section; detecting vibration of the wall at a minimum of two axial positions within said isolated sections; and detecting the reduction in vibration level of the wall at one or more of said axial positions in comparison to the remaining axial positions. The apparatus comprises a tool head to be inserted into the pressure tube, the tool head comprising a first end and a second and a clamping block m each of said ends. The clamping blocks are used to vibrationally isolate a section of the pressure tube located between said ends. The apparatus also comprises piezo-actuators operable to vibrate said pressure tube; and accelerometers used for measuring vibration of said pressure tube. |
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description | The present invention claims priority under 35 U.S.C. 119(e) to the provisional patent application filed on Jun. 24, 2013 and assigned application No. 61/838,692. This provisional patent application is incorporated in its entirety herein. The present invention applies to betavoltaic batteries having increased active area (e.g., surface area) to increase device efficiency by absorbing more beta particles. The direct conversion of radioisotope beta (electron) emissions into usable electrical power via beta emissions directly impinging on a semiconductor junction device was first proposed in the 1950's. Incident beta particles absorbed in a semiconductor create electron-hole-pairs (EHPs) which are accelerated by the built-in field to device terminals, and result in a current supplied to a load resistor. These devices are known as Direct Conversion Semiconductor Devices, Beta Cells, Betavoltaic Devices, Betavoltaic Batteries, Isotope Batteries etc. These direct conversion devices promise to deliver consistent long-term battery power for years and even decades. For this reason, many attempts have been made to commercialize such a device. However, in the hopes of achieving reasonable power levels, the radioisotope of choice often emitted unsafe amounts of high energy radiation that would either quickly degrade semiconductor device properties within the betavoltaic battery or the surrounding electronic devices powered by the battery. The radiated energy may also be harmful to operators in the vicinity of the battery. As a result of these disadvantages and in an effort to gain approval from nuclear regulatory agencies for these types of batteries, the choice for radioisotopes has been limited to low-energy beta (electron) emitting radioisotopes, such as nickel-63, promethium-147 or tritium. Due to the fact that promethium-147 is regulated more stringently, requires considerable shielding, and nickel-63 has a relatively low beta flux, tritium has emerged as a leading candidate for such a battery device. Tritium betavoltaic batteries, sometimes referred to as tritium betavoltaic devices or tritium direct conversion devices, have been promoted during the last thirty years. Tritium is a relatively benign radioisotope with low beta energy emission that can easily be shielded with as little as a thin sheet of paper. Tritium has a long track record in commercial use in illumination devices such as EXIT signs in commercial aircraft, stores, school buildings and theatres. It is also widely used in gun sights and watch dials, making it an ideal power source for the direct conversion devices. Given the low power and relatively large size of a typical tritium betavoltaic cell, it has been difficult to produce a device with meaningful power that is both cost-effective and space-efficient. Several attempts have been made to produce useful current from a tritium betavoltaic battery. For example, polycrystalline or amorphous semiconductor devices have been considered for tritium betavoltaic batteries based on the assumption that such devices would allow batteries to be fabricated at a reduced cost. It is assumed that these devices could be manufactured in a thin-film like fashion and that tritium could be embedded within the polycrystalline or amorphous devices. However, this approach is extremely inefficient (much less than 1%) with respect to the beta energy emissions entering the semiconductor. The main reason for this low semiconductor conversion efficiency is the high dark current or leakage current of the semiconductor that acts as a negative current. This high dark current competes with the betavoltaic current produced by collection of EHPs created via the tritium beta particles impinging on the semiconductor. In short, the polycrystalline and amorphous semiconductors have a high number of defects resulting in recombination centers for the EHPs, which in turn significantly reduce the betavoltaic current and lead to very low efficiency for the battery. The best results for tritium betavoltaics have been achieved with single crystal semiconductor devices. Recent attempts have involved single crystalline semiconductor devices with a tritium source such as a tritiated polymer, aerogel or tritiated metal hydride placed in direct contact with a semiconductor junction device. Single crystalline semiconductors have longer carrier lifetimes and fewer defects resulting in much lower dark currents. Though successful use of single-crystal betavoltaics has improved the power yielded from a single betavoltaic cell, reductions in cost and the volume utilized for a single cell needs to decrease to allow betavoltaics to compete in the market place. In addition to the above listed obstacles, the texturing of a direct conversion semiconductor device for the purpose of increasing the surface area exposed to beta radiation emission has been proposed several times in the past. For example, on page 282 of the book entitled “Polymers, Phosphors and Voltaics for Radioisotope Microbatteries” edited by K. Bower et al., the use of porous silicon and tritium inserted into porous silicon holes was proposed as a means of increasing the surface area of the semiconductor device by 20 to 50 times, in contrast to the original planar semiconductor surface area. Each of the following published patent applications and patents propose a method for increasing the surface area of the semiconductor by textured growth of the semiconductor or a post-growth texturing method: US Patent Application Publication 2004/0154656 US Patent Application Publication 2007/0080605 U.S. Pat. No. 7,250,323 U.S. Pat. No. 6,949,865 U.S. Pat. No. 7,939,986 U.S. Pat. No. 7,663,288 US Patent Application Publication 2011/0079791 US Patent Application Publication 2007/0080605 Central to the approaches of the above-listed patents and published patent applications is a belief that an increase in surface area exposed to radioisotope emissions will increase the power per unit volume of the direct conversion semiconductor device. The overall goal of this approach is to not only reduce the size of the direct conversion device but also to potentially reduce the cost associated with producing the equivalent surface area in a planar semiconductor device. The problem with such an approach arises from several competing factors. The incident power from candidate radioisotopes for betavoltaics (e.g. tritium, promethium-147, nickel-63) is quite small per unit area exposed, the dark current of the semiconductor device is a very significant factor in the overall efficiency of the device; this is especially problematic when tritium is utilized. If the dark current of a device is high, due to recombination or trapping defects in the semiconductor, then the efficiency will be especially low. For this reason, it is preferable to use single crystal semiconductors that maintain a seed or preferred orientation where device defects are minimized and the dark current is sufficiently low so that power can be produced efficiently. Unfortunately, alterations to the semiconductor junction's crystal structure, as proposed in the above-listed patents and published patent applications, risk increasing lattice defects, resulting in a high number of recombination centers for EHPs. Using the conventional processes (e.g. surface modification of single crystal betavoltaic junctions via etching/micromachining techniques in the above patents and published patent applications) typically results in creation of a direct conversion semiconductor device with a low open circuit voltage and reduced short circuit current resulting in a low overall efficiency. In addition, edge-effects associated with highly articulated surfaces contribute to generation of trapping and recombination centers leading to overall low efficiency. The present invention relates to a tritium direct conversion semiconductor device comprised of a III-V semiconductor single crystal grown by, in one embodiment, a molecular beam epitaxy (MBE) process or, in another embodiment, by a metal organic chemical vapor deposition (MOCVD) process. The invention comprises a device structure with both a low dark current and high efficiency for conversion of tritium's beta emissions into electrical power. It should be understood that the high efficiency and longevity (e.g. over 10 years) of the various device structure embodiments are suitable for use with other candidate radioisotopes for betavoltaic operations (e.g., promethium-147 and nickel-63). One embodiment of the present invention proposes the inclusion of novel structural features within an Indium Gallium Phosphide homojunction semiconductor or betavoltaic junction 8 (comprising individual stacked layers not illustrated) in conjunction with a tritiated metal hydride source 10, as illustrated in FIG. 1, for supplying power to a load 12. A substrate is not illustrated in FIG. 1, although those skilled in the art recognize that a substrate is typically present. Those novel structural features are shown in FIGS. 2, 3 and 6. The tritiated metal hydride source (e.g., scandium tritide, titanium tritide, palladium tritide, magnesium tritide, lithium tritide, or any combination thereof, etc.) is directly in contact with the homojunction semiconductor to generate electrical power at an efficiency of about 7.5% or higher with respect to the beta electrons impinging on the Indium Gallium Phosphide (InGaP) homojunction 8. InGaP is one of the larger band gap materials and has only recently been used in a tritium-based direct conversion battery. One embodiment uses a composition of the Indium Gallium Phosphide homojunction comprising In0.49 Ga0.51 P (referred to herein as InGaP). The band gap of this semiconductor is 1.9 eV and the materials production technology has been well developed by the solar cell industry. The technology also lends itself to high quality growth with a low density of lattice defects and low dark current characteristics. In addition, InGaP may be mass produced with a high yield due to its manufacturing process maturity, thus lowering the cost of tritium betavoltaic batteries based on InGaP. InGaP device structures can be grown by metal-organic-vapor-deposition (MOCVD) as is known by those skilled in the art. Other embodiments and physical layer structures are described and claimed in other patents and various other co-pending applications of the current assignee, e.g., U.S. Pat. Nos. 8,634,201 and 8,487,507, and co-pending patent application Ser. No. 13/925,736 (filed 24 Jun. 2013), Ser. No. 61/940,571 (filed 16 Feb. 2014) and Ser. No. 14/304,687 (filed Jun. 13, 2014). The description of embodiments related to a direct conversion semiconductor having an increased active (surface) area and the novel, the non-obvious features thereof and a method for fabricating same are illustrated in various figures of the present application. These features and structures allow more efficient conversion of tritium beta flux to electrical power than known in the prior art and can be applied to other embodiments and physical layer structures, such as those represented by the patents and applications referred to in the immediately-preceding paragraph. Unique to one embodiment of the present invention, enhanced surface area is achieved prior to deposition of semiconductor junction material. A betavoltaic junction or homojunction, such as an InGaP homojunction, is summarily grown conformally atop the patterned features that have been created in an underlying layer (a substrate for example) via methods known in the art (i.e. etching, deep reactive ion etching, nano-/micro-machining etc.) with attention to maintenance of crystal plane preferred orientation. This significant difference in design and processing from the prior art reduces potential contributions to dark current by decreasing defect structure population. Some very minor contributions to defect structures in the form of discrete departures from the optimal preferred orientations can occur; these departures can be caused by inaccuracies in the manufacturing techniques, and in some cases, caused by changes in radii at geometric edges for any chosen shapes. However, despite the occurrence of such non-optimal orientations, the overall performance of the semiconductor device reflects the fact that the majority of surfaces do exist with the optimal preferred orientation. Additionally, any resultant defect structures that occur can be mitigated or healed through annealing, etching, and other techniques known in the art, prior to growth of the III-V material, allowing for a higher quality betavoltaic junction as compared to prior art inventions. These techniques and features stand in stark contrast to the prior art where the betavoltaic junction's lattice is modified either during or after its formation to increase the device's surface area. Such modifications as described in the prior art also increase the junction's defect population and reduce overall efficiency. In the present invention the amount of active area of a semiconductor device is increased without a concomitant (or with only an insignificant increase) in the volume or dark current per unit area of that semiconductor, leading to an increased energy density. To that end there are several topographical structures and techniques discussed herein to increase the active area, decrease the semiconductor volume, or in general, increase the ratio of betavoltaic active area to betavoltaic volume. Central to this invention is the growth of III-V thin-film epitaxial layers conformally on top of a patterned or textured semiconductor substrate surface with articulated features such as pillars, trenches, triangles, pyramids, mounds, squares with rounded corners or other geometric shapes or features (e.g., shaped features) as shown in FIGS. 2, 3 and 6. The formation of these topographical 3D structures on the semiconductor substrate surface may be created with techniques such as, reactive ion etching, wet chemical etching, micro or nano machining, deep reactive ion etching, or utilizing other techniques known in the art that will create a smooth seed layer along the semiconductor substrate's crystal planes suitable for MBE or MOCVD growth of a overlying III-V betavoltaic structure, e.g. a p/n junction or an n/p junction depending on the doping of the substrate. These techniques, when combined with proper device design, create additional surface area while minimizing defects harmful to betavoltaic efficiency. The techniques and structures utilized keep the semiconductor volume relatively constant while increasing the surface area ratio of junction area formed on the patterned surface relative to that for a flat surface (e.g. up to 10 times the planar surface area or greater), and, by extension, also increasing the power and energy density of the betavoltaic device. MBE and MOCVD growth of III-V structures are mature processes that can create junctions with minimal defects allowing for high efficiency betavoltaic devices (See for example, Phys. Today issue of December 2012). FIGS. 2 and 3 illustrate two of the many possible approaches for achieving increased surface area, namely, pillars and trenches. Generally, any non-planar shape will increase the surface area. For both approaches, processing begins by utilizing deep reactive ion etching (DRIE) or other techniques known in the art to form geometric patterns (as depicted by FIGS. 2 and 3) on the bare semiconductor substrate (e.g. GaAs, Ge, GaP and Silicon, etc). The substrates may be N-doped or P-doped. In another embodiment it may be desired to grow an MBE/MOCVD foundational layer on the substrate. In one embodiment a thickness of the foundational layer is equivalent to the greatest depth to which a geometric feature will be etched. The foundational layer may be useful in instances where a derived benefit may be experienced by the semiconductor junction (e.g. high crystal quality for seed layer MOCVD growth). The derived benefit referred to here is the high quality crystal for the seed layer. Thus the lower defect population on the faces of the exemplary trenches or pillars of FIGS. 2 and 3 are important for achieving a high quality growth of the MBE/MOCVD growth (i.e., the betavoltaic junction). The array of geometric patterned shapes greatly increases the possible surface area that can subsequently receive deposition of semiconductor junction material via MBE/MOCVD processes, for example. After the DRIE process or another process for forming the patterns, the sample may be annealed and/or exposed to wet/dry etch to remove damage and/or contaminants and/or detritus left behind by the DRIE process, leaving a high quality surface for junction formation. As the layers of the junction are formed, their shape conforms to the shape of the underlying layer. In one embodiment, a preferred crystallographic planar orientation is maintained on all orthogonal surfaces of the topographically modified (i.e., shaped features formed therein) III-V semiconductor substrate. It should be noted in one embodiment that the parent substrates are initially procured with a (100) preferred orientation in plan; this (100) preferred orientation is known to the inventors to produce a seeding pattern for subsequent p/n junction growth that produces an optimal betavoltaic performance in the III-V semiconductor. Given a planar substrate surface as a starting point that possesses the (100) preferred orientation, subsequent semiconductor nucleation and growth on that surface produces p/n junctions that preserve the (100) seed orientation, as growth occurs along the [100] direction (orthogonal to the (100) plane). Given the symmetry of some cubic (zinc-blende) structures (e.g., GaAs), other diamond cubic structures (e.g. Ge, Si), and other similar materials, planes orthogonal to the (100) plane, namely (010) and (001) planes, are crystallographically identical to the (100) plane. Therefore, orthogonal surface modifications of the raw substrate that seek to produce the articulated 3-D structures (pillars, trenches) simply expose these other orthogonal (but crystallographically identical) planes. As illustrated in FIGS. 2 and 3, substrates are modified to produce extended orthogonal surface areas such that the (100) preferred seed orientation is still preserved throughout the articulated 3-D modified surface. The combination of the substrate's unique cubic symmetry in concert with the initial (100) seed layer orientation, therefore, allows for subsequent deposition (via nucleation and growth) of p/n junction materials that maintain the optimal (100) preferred orientation on all of the planar (orthogonal) surfaces produced by the exposed geometric faces in the articulated 3-D substrate. After formation of the shaped features the fabrication process continues with growing an n/p or p/n betavoltaic junction on the faces of the textured semiconductor substrate via MBE or MOCVD. Some betavoltaic junctions that may be grown are described in one or more of the commonly owned patents and patent applications identified above. The novel construction of these 3D topographical structures with each side having an identical orientation (e.g. (100) orientation) allows for uniform growth of the n/p or p/n junction on each face of the articulated surface. One embodiment uses a composition of the Indium Gallium Phosphide homojunction comprising In0.49 Ga0.51 P (subsequently referred to as InGaP) grown via MBE or MOCVD. The InGaP structure is grown conformally on faces of the textured surface of the underlying substrate. The band gap of this semiconductor junction is 1.9 eV and the materials production technology is well developed by the solar cell industry. The technology also lends itself to high quality growth with a low density of lattice defects and low dark current characteristics. In addition, InGaP may be mass produced with a high yield due to its manufacturing process maturity, thus lowering the cost of tritium betavoltaic batteries based on InGaP. InGaP device structures are normally grown by metal-organic-vapor-deposition (MOCVD) as is known by those skilled in the art. Additionally, the growth of InGaP on these articulated faces or shaped features can yield conformal betavoltaic structures with aspect ratios of up to 10:1 or greater, yielding even lower cost betavoltaic devices with high power outputs and energy densities. FIG. 4 illustrates exemplary individual layers of the n/p homojunction semiconductor 8, (note that if the substrate is doped p type then it is referred to as an n/p (n on p) homojunction and if the substrate is doped n type then it is referred to as a p/n homojunction (p on n)) comprising, from the bottom: a pGaAs substrate a p+InGaP layer (a back surface field (BSF) or minority carrier reflector) a pInGaP base layer an intrinsic InGaP layer (for preventing diffusion of dopants between the p-doped and n-doped layers) an nInGaP emitter layer an nInAlP layer (window layer closely lattice matched to the nInGaP emitter layer and the nGaAs cap layer that allows electrons to pass to the nGaAs cap layer and reflects holes back to the emitter) an nGaAs cap layer (may be highly doped in one embodiment) It should be noted that in other embodiments the dopant types may be reversed from those set forth above, additional layers may be added, and certain layers deleted from the layers presented in FIG. 4. If the dopant types are reversed from those set forth above, the structure may be referred to as a p/n homojunction semiconductor. Additionally, any of the substrates and/or structures described in commonly-owned patents and patent applications, including those listed herein, may be grown on the 3D high aspect ratio or high surface area modified (i.e., including shaped features formed therein) substrates as described herein. There are several features of this n/p structure that allow efficient betavoltaic energy conversion: (a) High quality, large band gap semiconductor junction resulting in a highly efficient device; (b) Back-surface field (BSF) reflective layer comprising a highly doped p+InGaP layer (can also be created by p-type InAlP, InAlGaP or ZnSe); (c) A lattice-matched n-type InAlP window layer to reflect holes back to the emitter leading to a low dark current (can also be created with a highly doped n+pseudomorphic material, InAlGaP, ZnSe, AlAs or AlAsP); (d) A GaAs cap layer of about a few hundred angstroms or less covering the top surface; and (e) a 1000 to 3000 Å layer of intrinsic InGaP to act as a buffer to diffusion of the p type dopant (usually Zn) into the n-type emitter region. The intrinsic layer may also be less than 100 Å. Each of these features contributes to the achievement of the low dark currents required for efficient betavoltaic energy conversion. The novel lattice-matched InAlP window layer prevents the formation of dislocations at the InAlP-InGaP interface, which would increase the dark current. The GaAs cap layer keeps the InAlP layer from oxidizing, the absence of which could introduce defects for EHP recombination at the InAlP-InGaP region. This cap layer, therefore augments hole reflection at that interface. The GaAs cap layer does not absorb a significant percentage of the beta flux, and therefore the small absorption can be tolerated. It should be noted that the cap layer can be made out of other III-V materials or combinations of III-V materials that can function in a similar capacity. The GaAs Cap layer acts as a current collector for the betavoltaic junction serving yet another purpose. Tritium beta particle penetration in semiconductors is less than about one micron. Thus, it is clear that the emitter, window and GaAs cap layers need to be very thin so that most of the beta particle absorption occurs in the high field region in the depletion layer (with respect to FIG. 4, the intrinsic InGaP layer or in another embodiment a material region between a p-doped and an n-doped region). As shown in FIG. 4, ohmic contacts 40 and 42 for both polarities can be made to the GaAs substrate and the GaAs cap layer. The betavoltaic device with shaped features is completed by conformally coating a hydride-capable metal (e.g titanium, magnesium, scandium, lithium, palladium etc.) via methods known in the art (e.g. evaporation, sputtering, electro-deposition, atomic layer deposition, chemical vapor deposition etc.). This conformal coating is directly deposited onto the faces of the 3D modified surfaces that now comprise the betavoltaic structure. Tritium may also be stored as part of a polymer or aerogel that is then conformally coated or deposited onto the faces of the 3D betavoltaic structure. Other radioisotopes may also be deposited onto the 3D betavoltaic structures (e.g. nickel-63, promethium-147, phosphorus-33). In the case of hydride-capable metals conformally coated onto the betavoltaic structure, the metal tritide is formed by exposure to tritium gas at pressures approximately ranging from less than 0.01 to greater than 20 Bar and temperatures ranging from approximately room temperature to 600° C. for durations ranging minutes to days. Thicknesses of the metal tritide layer are typically on the order of microns and can be as thin as 50-100 nanometers. A cap layer of palladium ranging from approximately 1.0 nanometer to 500.0 nanometers may be deposited over a scandium, titanium, magnesium, lithium, or other suitable metal in order to reduce the tritium loading temperature and stabilize the tritium within the metal matrix after the tritide has been formed. The palladium cap layer functions primarily as a catalyst and serves to provide for an expedited rate of reaction for inducing the process of tritiation; palladium has an additional benefit for the tritiation process in that it can facilitate tritium loading of a metal tritide at significantly lower temperatures and pressures compared to processing efforts conducted in the absence of palladium. This subsequent increase in the kinetics of the tritiation process induced by the palladium cap layer does not alter the ultimate functionality of the betavoltaic cell, and it is usually deposited directly upon un-passivated surfaces (surfaces containing no oxide barriers to tritiation) of metal hydride storage layers (e.g. scandium, titanium, magnesium, lithium, or other suitable metal). The palladium layer is typically laid down in a vacuum/inert gas atmosphere process, in order to eliminate oxygen contamination and is deposited via any of the metal deposition techniques described elsewhere herein. It should also be noted that the metal tritide layer may also be formed via an in-situ evaporation of the metal in the presence of gaseous tritium. In some embodiments, the metal tritide is conductive (e.g. titanium tritide, scandium tritide) and can serve as both a beta emitting source and an ohmic contact. In cases where the tritide is an insulator, a regular metal point contact to the GaAs cap layer may be utilized as illustrated in FIG. 4. In one embodiment with pillar formations of FIG. 3, the junction area formed on the top of the pillars, the sides of the pillars and in between pillars by MOCVD can be characterized by a large aspect ratio. In particular, the junction area is increased by a factor of {1+[4hs/(r+s)2]} where dimensions h, s and r are defined in FIG. 3 and for cases where r=s and r=s/2. FIG. 5 shows the result of varying these values. The increased aspect ratio gives rise to increased surface area and hence an increased value of short circuit current. For embodiments with trench formations as described by FIG. 2, the junction is formed on top of the ridges, on the sides of the ridges and in between the ridges. Similarly, one finds that if the surface is modified with trenches, the surface area can increase by a factor of {1+[2h/(r+s)]}, again where the dimensions h, s, and r are identified in FIG. 2. As in the case of pillars, the increased aspect ratio leads to the possibility of increased values of short circuit current. This is due to the high quality of MBE or MOCVD growth of the n/p or p/n structure that can be made on the (100) orientation of all faces of the pillars or trenches. FIG. 5 illustrates the resulting aspect ratio value as a function of h for specified values of s and r for trench and pillar features. FIG. 6 illustrates the various layers of FIG. 4 from a perspective where the shaped features can be clearly seen. For example, looking down or along the trenches 40 of FIG. 2 one would see the layered structures with the shaped features of FIG. 6. In all embodiments of the present invention, the voltage and current may be scaled up via stacking of the aforementioned betavoltaic cells with high aspect ratio extended surface area. Betavoltaic cell layers may be stacked vertically or arranged horizontally and configured electrically in series or parallel. Electrical connection can be established by utilizing through-vias as power lead contacts across betavoltaic cell layers, by using current-channeling interposers (e.g. flexible circuit cards) in between betavoltaic cells or groups of cells, or by other methods common in the art. It should be noted that varying stacking configurations produce varying voltage and current outputs from the betavoltaic composite device. An additional approach to increasing the power/energy density of stacked high aspect ratio extended surface area betavoltaic cells is to thin the back surfaces of the betavoltaic semiconductor substrates by etching, polishing, grinding, epitaxial-lift-off and/or other lapidary techniques known in the art. It should be noted that typical starting points for GaAs substrate thicknesses are 625 microns, but can be thinned using the aforementioned methods. Similarly, Germanium substrates are typically 175 microns thick but may be thinned using methods known in the art. The referenced aspect ratio of the patterned features of shaped features is defined as the ratio of the opening or cut to the depth of the opening or cut. Certain embodiments of the invention exhibit an aspect ratio of at least 2:1 to as great as 500:1. While certain embodiments of the present invention have been shown and described herein, such embodiments are provided by way of example only. Numerous variations, changes and substitutions will occur to those of skill in the art without departing from the invention herein. Accordingly, it is intended that the invention be limited only by the spirit and scope of the appended claims. |
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abstract | The object of the present invention is to provide a satisfactory image at a desired imaging timing by implementing grid movement control according to the time response characteristics of the radiation generation function and a decrease in time delay from an imaging request to actual irradiation. In order to achieve this object, a control device controls the actual irradiation instruction timing for an irradiation device on the basis of a pre-irradiation delay time as a time between an instruction and irradiation of actual irradiation of the irradiation device. |
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059404612 | abstract | To provide a reactor core for light water cooled reactor, a fuel assembly and a control rod intended for Pu multi-recycling at a breeding ratio of about 1.0, or 1.0 or more while keeping the economical or safety performance to the same level as in BWR now under operation, that is, while minimizing the change for the incore structures and maintaining the void coefficient negative. A reactor core for a light water cooled reactor having an effective fuel-to-water volume ratio of 0.1 to 0.6 by the combination of a dense lattice fuel assembly constituted of fuel rods formed by adding Pu to degraded uranium, natural uranium, depleted uranium or low concentrated uranium, coolants at high void fraction of 45% to 70% and a cluster-type, Y-type or cruciform control rod. |
claims | 1. A method of making a zirconium based alloy article which exhibits one of improved corrosion resistance and improved creep resistance for use in an elevated temperature environment of a nuclear reactor, comprising the steps:(a) combining0.4 to 1.5 weight percent niobium,0.4 to 0.8 weight percent tin,0.05 to 0.3 weight percent iron,0.0 to 0.5 weight percent chromium, andthe balance at least 97 weight percent zirconium including impurities to provide an alloy mixture;(b) melting the alloy mixture to produce a melted alloy material;(c) forging the melted alloy material to produce a forged alloy material;(d) quenching the forged alloy material to produce a quenched alloy material;(e) rolling the quenched alloy material to produce a rolled alloy material;(f) annealing the rolled alloy material to produce an annealed alloy material;(g) following completion of steps (b), (c), (d), (e) and (f), subjecting the annealed alloy material to a final heat treatment selected to provide a zirconium-based alloy exhibiting one of improved corrosion resistance and improved creep resistance,wherein for providing the zirconium-based alloy exhibiting improved corrosion resistance, the annealed alloy material is subjected to a final heat treatment of partial recrystallization to produce an amount of recrystallization from about 15% to 20% with the remainder being stress relief annealed, andwherein for providing the zirconium-based alloy exhibiting improved creep resistance, the annealed alloy material is subjected to a final heat treatment of partial recrystallization to produce an amount of recrystallization from about 80% to 95% recrystallization with the remainder being stress relief annealed. 2. The method of making the zirconium based alloy of claim 1, wherein the alloy has a composition of about:0.4 to 1.5 weight percent niobium,0.6 to 0.7 weight percent tin,0.05 to 0.3 weight percent iron,the balance at least 97 weight percent zirconium, including impurities. 3. The method of making the zirconium based alloy of claim 1, wherein the alloy has a composition of about:0.4 to 1.5 weight percent niobium,0.61 to 0.69 weight percent tin,0.05 to 0.3 weight percent iron,the balance at least 97 weight percent zirconium, including impurities. 4. The method of making the zirconium alloy of claim 1, wherein the alloy further comprises:0.05 to 0.5 weight percent chromium. |
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041939530 | abstract | A hydrosol containing, in nitrate form, a fuel or fuel-and-breeder material which is projected horizontally in the form of droplets into a gas phase containing gaseous ammonia and allowed to fall in a drip-casting column into a precipitation bath containing ammonium hydroxide. In the gas phase, the droplets are hardened just enough to prevent their deformation upon penetrating into the precipitation bath where the hardening is completed. A falling height of 5 cm is suitable. The granules are washed free of ammonium nitrate, then dried, and then sintered. The heavy metal content in the hydrosol is between 1.5 and 3 moles per liter, and the pH value of the precipitation bath is between 8 and 9. The hydrosol contains the heavy metal in oxide form and the process can be used with a thorium oxide hydrosol or a hydrosol that, in addition to thorium oxide, contains the oxide of hexavalent uranium, in the latter case the hexavalent uranium being present in a proportion up to 25% by weight of the total heavy metal. The process is also applicable to producing kernels of mixed thorium and plutonium oxides. In the case of uranium-containing granules, the sintering step is carried out in a reducing atmosphere to convert the uranium to the tetravalent state. |
claims | 1. A hazardous material storage repository, comprising:a drillhole extending into the Earth and comprising an entry at least proximate a terranean surface, the drillhole comprising a substantially vertical drillhole portion, a transition drillhole portion coupled to the substantially vertical drillhole portion, and a hazardous material storage drillhole portion coupled to the transition drillhole portion, at least one of the transition drillhole portion or the hazardous material storage drillhole portion comprising an isolation drillhole portion that is directed vertically toward the terranean surface and away from an intersection between the substantially vertical drillhole portion and the transition drillhole portion;a storage canister positioned in the hazardous material storage drillhole portion, the storage canister sized to fit from the drillhole entry through the substantially vertical drillhole portion, the transition drillhole portion, and into the hazardous material storage drillhole portion of the drillhole, the storage canister comprising an inner cavity sized enclose hazardous material; anda seal positioned in the drillhole, the seal isolating the hazardous material storage drillhole portion of the drillhole from the entry of the drillhole. 2. The hazardous material storage repository of claim 1, wherein the isolation drillhole portion comprises a vertically inclined drillhole portion that comprises a proximate end coupled to the transition drillhole portion at a first depth and a distal end opposite the proximate end at a second depth shallower than the first depth. 3. The hazardous material storage repository of claim 2, wherein the vertically inclined drillhole portion comprises the hazardous material storage drillhole portion. 4. The hazardous material storage repository of claim 2, wherein an inclination angle of the vertically inclined drillhole portion is determined based at least in part on a distance associated with a disturbed zone of a geologic formation that surrounds the vertically inclined drillhole portion and a length of a distance tangent to a lowest portion of the storage canister and the substantially vertical drillhole portion. 5. The hazardous material storage repository of claim 4, wherein the distance associated with the disturbed zone of the geologic formation comprises a distance between an outer circumference of the disturbed zone and a radial centerline of the vertically inclined drillhole portion. 6. The hazardous material storage repository of claim 4, wherein the inclination angle is about 3 degrees. 7. The hazardous material storage repository of claim 1, wherein the isolation drillhole portion comprises a J-section drillhole portion coupled between the substantially vertical drillhole portion and the hazardous material storage drillhole portion. 8. The hazardous material storage repository of claim 7, wherein the J-section drillhole portion comprises the transition drillhole portion. 9. The hazardous material storage repository of claim 7, wherein the hazardous material storage drillhole portion comprises at least one of a substantially horizontal drillhole portion or a vertically inclined drillhole portion. 10. The hazardous material storage repository of claim 1, wherein the isolation drillhole portion comprises a vertically undulating drillhole portion coupled to the transition drillhole portion. 11. The hazardous material storage repository of claim 10, wherein the transition drillhole portion comprises a curved drillhole portion between the substantially vertical drillhole portion and the vertically undulating drillhole portion. 12. The hazardous material storage repository of claim 1, wherein the hazardous material storage drillhole portion is located within or below a barrier layer that comprises at least one of a shale formation layer, a salt formation layer, or other impermeable formation layer. 13. The hazardous material storage repository of claim 12, wherein the hazardous material storage drillhole portion is vertically isolated, by the barrier layer, from a subterranean zone that comprises mobile water. 14. The hazardous material storage repository of claim 12, wherein the hazardous material storage drillhole portion is formed below the barrier layer and is vertically isolated from the subterranean zone that comprises mobile water by the barrier layer. 15. The hazardous material storage repository of claim 12, wherein the hazardous material storage drillhole portion is formed within the barrier layer, and is vertically isolated from the subterranean zone that comprises mobile water by at least a portion of the barrier layer. 16. The hazardous material storage repository of claim 12, wherein the barrier layer comprises a permeability of less than about 0.01 millidarcys. 17. The hazardous material storage repository of claim 12, wherein the barrier layer comprises a brittleness of less than about 10 MPa, where brittleness comprises a ratio of compressive stress of the barrier layer to tensile strength of the barrier layer. 18. The hazardous material storage repository of claim 12, wherein the barrier layer comprises a thickness proximate the hazardous material storage drillhole portion of at least about 100 feet. 19. The hazardous material storage repository of claim 12, wherein the barrier layer comprises a thickness proximate the hazardous material storage drillhole portion that inhibits diffusion of the hazardous material that escapes the storage canister through the barrier layer for an amount of time that is based on a half-life of the hazardous material. 20. The hazardous material storage repository of claim 12, wherein the barrier layer comprises about 20 to 30% weight by volume of clay or organic matter. 21. The hazardous material storage repository of claim 12, wherein the barrier layer comprises an impermeable layer. 22. The hazardous material storage repository of claim 12, wherein the barrier layer comprises a leakage barrier defined by a time constant for leakage of the hazardous material of 10,000 years or more. 23. The hazardous material storage repository of claim 12, wherein the barrier layer comprises a hydrocarbon or carbon dioxide bearing formation. 24. The hazardous material storage repository of claim 1, wherein the hazardous material comprises spent nuclear fuel. 25. The hazardous material storage repository of claim 1, further comprising at least one casing assembly that extends from at or proximate the terranean surface, through the drillhole, and into the hazardous material storage drillhole portion. 26. The hazardous material storage repository of claim 1, wherein the storage canister comprises a connecting portion configured to couple to at least one of a downhole tool string or another storage canister. 27. The hazardous material storage repository of claim 1, wherein the isolation drillhole portion comprises a spiral drillhole. 28. The hazardous material storage repository of claim 1, wherein the isolation drillhole portion comprises a specified geometry independent of a stress state of a rock formation into which the isolation drillhole portion is formed. 29. A hazardous material storage repository, comprising:a drillhole extending into the Earth and comprising an entry at least proximate a terranean surface, the drillhole comprising a substantially vertical drillhole portion, a transition drillhole portion coupled to the substantially vertical drillhole portion, and a hazardous material storage drillhole portion coupled to the transition drillhole portion, the hazardous material storage drillhole portion located below a self-healing geological formation, the hazardous material storage drillhole portion vertically isolated, by the self-healing geological formation, from a subterranean zone that comprises mobile water, at least one of the transition drillhole portion or the hazardous material storage drillhole portion comprising an isolation drillhole portion that is directed vertically toward the terranean surface and away from an intersection between the substantially vertical drillhole portion and the transition drillhole portion;a storage canister positioned in the hazardous material storage drillhole portion, the storage canister sized to fit from the drillhole entry through the substantially vertical drillhole portion, the transition drillhole portion, and into the hazardous material storage drillhole portion of the drillhole, the storage canister comprising an inner cavity sized enclose hazardous material; anda seal positioned in the drillhole, the seal isolating the hazardous material storage drillhole portion of the drillhole from the entry of the drillhole. 30. The hazardous material storage repository of claim 29, wherein the self-healing geologic formation comprises at least one of shale salt, clay, or dolomite. |
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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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048448389 | abstract | A method of treatment of a radioactive liquid waste containing fission products and a thermally decomposable sodium compound which comprises heating the liquid waste to convert the sodium compound into oxides of sodium, converting the oxides into sodium, hydroxide, reacting the sodium hydroxide with an alcohol to form a sodium alcoholate, separating the sodium alcoholate from an impurity residue essentially comprising fission products, decomposing the separated sodium alcoholate to form sodium hydroxide and recovering the sodium hydroxide. |
058898344 | abstract | The accuracy of radiation therapy is enhanced with the blade collimator of the claimed invention. Particularly, this collimator includes a support device for supporting at least two sets of radiation attenuating blades in a single plane or layer. The blades are reciprocally movable within this support. Shaping of the therapy may be increased by constructing the blades to include a varying width increasing from the central blade toward the exterior blades. |
summary | ||
description | The present invention relates to a water jet peening apparatus and a water jet peening method used to repair a nozzle provided in a nuclear reactor vessel and the vicinity thereof. For example, a nuclear power plant that includes a pressurized water reactor (PWR) uses light water as a nuclear reactor coolant and a neutron moderator while keeping the light water as high-temperature and high-pressure water which is not boiled throughout a reactor core, sends the high-temperature and high-pressure water to a steam generator so as to generate a steam by a heat exchange operation, and sends the steam to a turbine generator so as to generate electric power. In such a nuclear power plant, there is a need to periodically inspect various structures of the pressurized water reactor in order to ensure sufficient safety or reliability. Then, when a problem is found after various inspections, a necessary portion involved with the problem is repaired. For example, in the pressurized water reactor, a nuclear reactor vessel body is provided with a plurality of instrumentation nozzles penetrating a lower end plate. Further, each of the instrumentation nozzles is formed so that an in-core instrumentation guide pipe is fixed to the upper end thereof inside the reactor and a conduit tube is connected to the lower end thereof outside the reactor. Then, a neutron flux detector capable of measuring a neutron flux is insertable from the instrumentation nozzle to a reactor core (a fuel assembly) through the in-core instrumentation guide pipe by using the conduit tube. The instrumentation nozzle is formed in a manner such that an in-core instrumentation cylinder is welded while being fitted into an attachment hole of a nuclear reactor vessel body. For that reason, there is a possibility that a tensile stress may remain in the in-core instrumentation cylinder, the welding portion of the in-core instrumentation cylinder, and the vicinity thereof. Thus, there is an increase in the possibility of stress corrosion cracking due to the long-term use. Here, as the related art, a water jet peening technique is known which prevents stress corrosion cracking by solving a residual tensile stress of a surface by a residual compressive stress. In the water jet peening operation, high-pressure water including cavitation air bubbles is jetted to a surface of a metal member under water so as to solve a residual tensile stress of the surface of the metal member by a residual compressive stress. As such a water jet peening apparatus, for example, an example is disclosed in Patent Literature 1 as below. Patent Literature 1: JP 2006-201141 A Incidentally, in a case where the water jet peening operation is performed on the inner surface of the instrumentation nozzle (the in-core instrumentation cylinder), a neutron flux detector which is disposed inside the instrumentation nozzle is extracted by a predetermined length to the outside (the lower portion) through a conduit tube along with a thimble tube. However, when the water jet peening operation is performed on the inner surface of the in-core instrumentation cylinder, the pressure of the in-core instrumentation cylinder increases due to the jetted high-pressure water, and a neutron flux detector is further pressed by the water pressure. Accordingly, a problem arises in that the thimble tube moves so as to be popped out to a storage chamber. The invention is made to solve the above-described problems, and an object thereof is to provide a water jet peening apparatus and a water jet peening method capable of improving the safety of an operation by preventing a thimble tube from being popped out due to a water jet peening operation. According to an aspect of the present invention, a water jet peening apparatus includes: a clamping cylinder which is able to be disposed at the outer peripheral side of an instrumentation nozzle with a predetermined gap therebetween; a clamping mechanism which is able to fix the clamping cylinder to the instrumentation nozzle; a positioning member that has a cylindrical shape, is provided inside the clamping cylinder, and is positioned to a position adjacent to the upper end of the instrumentation nozzle; an inner surface WJP nozzle which is movable upward and downward inside the positioning member; and a drainage hole which radially penetrates the positioning member. Accordingly, the positioning member is provided at a position adjacent to the upper end of the instrumentation nozzle inside the clamping cylinder, and the drainage hole is provided so as to radially penetrate the positioning member. Accordingly, when the high-pressure water including cavitation air bubbles is jetted from the inner surface WJP nozzle to the inner surface of the instrumentation nozzle under water, the residual tensile stress of the inner surface of the instrumentation nozzle is solved by the residual compressive stress, and the jetted high-pressure water is discharged from the drainage hole of the positioning member while not substantially giving any action on the neutron flux detector. Accordingly, it is possible to improve the safety of the operation by preventing the thimble tube from being popped out due to the water jet peening operation. Advantageously, in the water jet peening apparatus, the positioning member includes a cylindrical portion and a flange portion provided at the lower end of the cylindrical portion, and a plurality of the drainage holes is provided in the flange portion at the same interval in the circumferential direction. Accordingly, since the plurality of drainage holes is provided in the flange portion of the positioning member at the same interval in the circumferential direction, the high-pressure water which is jetted from the inner surface WJP nozzle is highly efficiently discharged from the plurality of drainage holes, and hence the draining performance may be improved. Advantageously, in the water jet peening apparatus, the positioning member is positioned to a position contacting the upper end of the instrumentation nozzle. Accordingly, since the positioning member is positioned to a position contacting the upper end of the instrumentation nozzle, the positioning member and the instrumentation nozzle are disposed without any gap therebetween, and hence the high-pressure water which is jetted from the inner surface WJP nozzle may be appropriately led to the plurality of drainage holes. Advantageously, in the water jet peening apparatus, a detection device is provided which detects a state where the positioning member is positioned to a predetermined position of the instrumentation nozzle. Accordingly, since the detection device detects a state where the positioning member is positioned to a predetermined position of the instrumentation nozzle, the water jet peening apparatus may be highly precisely positioned to the instrumentation nozzle. Advantageously, in the water jet peening apparatus, the detection device includes a detection rod which is supported by the clamping cylinder so as to be movable upward and downward and of which the lower end directly or indirectly is able to contact the upper end of the instrumentation nozzle and a detector which detects the up and down position of the detection rod. Accordingly, when the lower end of the detection rod supported by the clamping cylinder so as to be movable upward and downward directly or indirectly contacts the upper end of the instrumentation nozzle, the detector detects the position of the detection rod, and hence the water jet peening apparatus may be highly precisely positioned to the instrumentation nozzle with a simple configuration. Advantageously, in the water jet peening apparatus, the positioning member includes an operation piece which is supported by the clamping cylinder so as to be movable upward and downward and able to come into surface-contact with the upper end surface of the instrumentation nozzle, and the detection rod is movable upward and downward through the operation piece. Accordingly, when the operation piece of the positioning member comes into surface-contact with the upper end surface of the instrumentation nozzle, the detection rod moves, and the movement of the positioning member to a predetermined position is highly precisely detected. Accordingly, the water jet peening apparatus may be highly precisely positioned to the instrumentation nozzle. Advantageously, in the water jet peening apparatus, the detection rod is disposed outside the positioning member. Accordingly, the inner surface WJP nozzle moves upward and downward inside the positioning member, and the detection rod is disposed outside the positioning member. Accordingly, the detection rod does not disturb the upward and downward movement of the inner surface WJP nozzle, and hence the stable water jet peening operation may be performed. Advantageously, in the water jet peening apparatus, one end of a conduit tube is connected to the instrumentation nozzle, and the other end thereof extends to a monitoring chamber so as to be connected thereto and a thimble tube having a neutron flux detector attached to the front end thereof is inserted from the other end of the conduit tube, and is insertable into a nuclear reactor vessel through the instrumentation nozzle, and a fixing device is provided which does not allow the end of the thimble tube drawn from the other end of the conduit tube to be movable with respect to the conduit tube. Accordingly, since the fixing device immovably fixes the thimble tube drawn from the other end of the conduit tube, it is possible to prevent the thimble tube from being popped out due to the water jet peening operation. Advantageously, in the water jet peening apparatus, the fixing device includes a first fixing jig which is fixed to the conduit tube extending to the monitoring chamber, a second fixing jig which is fixed to the end of the thimble tube drawn from the other end of the conduit tube, and a connection member which suppresses the separation of the first fixing jig and the second fixing jig. Accordingly, it is possible to easily prevent the thimble tube from being popped out due to the water jet peening operation with a simple configuration. Advantageously, in the water jet peening apparatus, a monitoring device is provided which monitors the fixed state of the thimble tube and the conduit tube. Accordingly, since the fixed state of the thimble tube and the conduit tube is monitored by the monitoring device, it is possible to prevent the thimble tube from being popped out due to the water jet peening operation. According to another aspect of the present invention, a water jet peening method includes: disposing a clamping cylinder at the outer peripheral side of an instrumentation nozzle with a predetermined gap therebetween; fixing a positioning member provided in the clamping cylinder to the instrumentation nozzle at a position adjacent to the upper end of the instrumentation nozzle; moving an inner surface WJP nozzle downward to the instrumentation nozzle through the clamping cylinder; jetting high-pressure water to an inner surface of the instrumentation nozzle by moving the inner surface WJP nozzle downward in a rotation state while the high-pressure water including cavitation air bubbles is jetted from the inner surface WJP nozzle; and discharging the high-pressure water jetted from the inner surface WJP nozzle to the outside from a drainage hole provided in the positioning member. Accordingly, when the high-pressure water including cavitation air bubbles is jetted from the inner surface WJP nozzle to the inner surface of the instrumentation nozzle under water, the residual tensile stress of the inner surface of the instrumentation nozzle is solved by the residual compressive stress, and the jetted high-pressure water is discharged from the drainage hole of the positioning member while not substantially giving any action on the neutron flux detector. Accordingly, it is possible to improve the safety of the operation by preventing the thimble tube from being popped out due to the water jet peening operation. Advantageously, in the water jet peening method, the high-pressure water is jetted from the inner surface WJP nozzle while a thimble tube drawn to the outside from the instrumentation nozzle through a conduit tube is not movable. Accordingly, it is possible to prevent the thimble tube from being popped out due to the water jet peening operation. Advantageously, in the water jet peening apparatus, the fixed state of the thimble tube and the conduit tube is monitored when the high-pressure water is jetted from the inner surface WJP nozzle to the inner surface of the instrumentation nozzle. Accordingly, it is possible to prevent the thimble tube from being popped out due to the water jet peening operation. According to the water jet peening apparatus and the water jet peening method of the invention, the positioning member is provided so as to be positioned to a position adjacent to the upper end of the instrumentation nozzle inside the clamping cylinder and the drainage hole is provided so as to radially penetrate the positioning member. Thus, it is possible to improve the safety of the operation by preventing the thimble tube from being popped out due to the water jet peening operation. Hereinafter, preferred embodiments of a water jet peening apparatus and a water jet peening method according to the invention will be described in detail with reference to the accompanying drawings. Furthermore, the invention is not limited to the embodiment. Further, when a plurality of embodiments is provided, the embodiments may be combined with one another. FIG. 11 is a schematic configuration diagram of a nuclear power plant, FIG. 12 is a longitudinal sectional view illustrating a pressurized water reactor, and FIG. 13 is a cross-sectional view illustrating an instrumentation nozzle of a nuclear reactor vessel. A nuclear reactor of the first embodiment is a pressurized water reactor that uses light water as a nuclear reactor coolant and a neutron moderator while keeping the light water as high-temperature and high-pressure water which is not boiled throughout a reactor core, sends the high-temperature and high-pressure water to a steam generator so as to generate a steam by a heat exchange operation, and sends the steam to a turbine generator so as to generate electric power. In a nuclear power plant that includes the pressurized water reactor of the first embodiment, as illustrated in FIG. 11, a containment 11 accommodates a pressurized water reactor 12 and a steam generator 13 therein. Here, the pressurized water reactor 12 and the steam generator 13 are connected to a high-temperature-side supply pipe 14 through a low-temperature-side supply pipe 15, the high-temperature-side supply pipe 14 is provided with a pressurizer 16, and the low-temperature-side supply pipe 15 is provided with a primary cooling water pump 17. In this case, light water is used as a moderator and primary cooling water (coolant), and a primary cooling system is controlled at a high-pressure state of about 150 to 160 atm by the pressurizer 16 in order to prevent the primary cooling water from being boiled in the reactor core. Accordingly, in the pressurized water reactor 12, the light water as the primary cooling water is heated by low-enriched uranium or MOX as a fuel (an atomic fuel), and the high-temperature primary cooling water is sent to the steam generator 13 through the high-temperature-side supply pipe 14 while being maintained at a predetermined high pressure by the pressurizer 16. In the steam generator 13, the primary cooling water which is cooled by a heat exchange operation between the high-temperature and high-pressure primary cooling water and the secondary cooling water is returned to the pressurized water reactor 12 through the low-temperature-side supply pipe 15. The steam generator 13 is connected to a steam turbine 32 through a pipe 31 that supplies the heated secondary cooling water, that is, steam, and the pipe 31 is provided with a main steam isolation valve 33. The steam turbine 32 includes a high-pressure turbine 34 and a low-pressure turbine 35, and is connected to a generator (a generation device) 36. Further, a moisture separation heater 37 is provided between the high-pressure turbine 34 and the low-pressure turbine 35. Here, a cooling water branch pipe 38 which is branched from the pipe 31 is connected to the moisture separation heater 37, the high-pressure turbine 34 and the moisture separation heater 37 are connected to each other by a low-temperature reheating pipe 39, and the moisture separation heater 37 and the low-pressure turbine 35 are connected to each other by a high-temperature reheating pipe 40. Further, the low-pressure turbine 35 of the steam turbine 32 includes a condenser 41. Here, the condenser 41 is connected to a turbine bypass pipe 43 which extends from the pipe 31 and includes a bypass valve 42, and is connected to a water intake pipe 44 and a drainage pipe 45 which supply and discharge the cooling water (for example, sea water). The water intake pipe 44 includes a circulation water pump 46, and the other end thereof is disposed under the sea along with the drainage pipe 45. Then, the condenser 41 is connected to a pipe 47, a condensate pump 48, a grand condenser 49, a condensed water desalting device 50, a condensate booster pump 51, and a low-pressure feed water heater 52. Further, the pipe 47 is connected to a deaerator 53, and is provided with a water feeding pump 54, a high-pressure feed water heater 55, and a water feeding control valve 56. Accordingly, in the steam generator 13, the steam which is generated by the heat exchange operation with respect to the high-pressure and high-temperature primary cooling water is sent to the steam turbine 32 (from the high-pressure turbine 34 to the low-pressure turbine 35) through the pipe 31. Then, the steam turbine 32 is driven by the steam so that the generator 36 generates electric power. At this time, the steam which is sent from the steam generator 13 is used to drive the high-pressure turbine 34, passes through the moisture separation heater 37 so that the steam is heated while a moisture contained in the steam is removed, and is used to drive the low-pressure turbine 35. Then, the steam having been used to drive the steam turbine 32 is cooled into condensed water by the sea water in the condenser 41, and is returned to the steam generator 13 through the grand condenser 49, the condensed water desalting device 50, the low-pressure feed water heater 52, the deaerator 53, the high-pressure feed water heater 55, and the like. In the pressurized water reactor 12 of the nuclear power plant with such a configuration, as illustrated in FIG. 12, a nuclear reactor vessel 61 includes a nuclear reactor vessel body 62 and a nuclear reactor vessel head (an upper end plate) 63 attached to the upper portion thereof so that an in-core structure is inserted thereinto, and the nuclear reactor vessel head 63 is fixed to the nuclear reactor vessel body 62 by a plurality of stud bolts 64 and a plurality of nuts 65 so as to be opened and closed. The nuclear reactor vessel body 62 has a cylindrical shape of which the upper portion can be opened by the separation of the nuclear reactor vessel head 63 and the lower portion is formed in a semi-spherical shape while being closed by a lower end plate 66. Then, the upper portion of the nuclear reactor vessel body 62 is provided with an inlet nozzle (an entrance nozzle) 67 which supplies the light water (coolant) as the primary cooling water and an outlet nozzle (an exist nozzle) 68 which discharges the light water. Further, the nuclear reactor vessel body 62 is provided with a water injection nozzle (a water injection nozzle) (not illustrated) separately from the inlet nozzle 67 and the outlet nozzle 68. In the inside of the nuclear reactor vessel body 62, an upper core support plate 69 is fixed to a portion above the inlet nozzle 67 and the outlet nozzle 68 and a lower core support plate 70 is fixed so as to be located in the vicinity of the lower end plate 66. The upper core support plate 69 and the lower core support plate 70 are formed in a disk shape and are provided with a plurality of flow holes (not illustrated). Then, the upper core support plate 69 is connected to an upper core plate 72 provided with a plurality of flow holes (not illustrated) below through a plurality of reactor core support rods 71. A core barrel 73 which has a cylindrical shape is disposed inside the nuclear reactor vessel body 62 with a predetermined gap with respect to the inner wall surface. Further, the upper portion of the core barrel 73 is connected to the upper core plate 72, and the lower portion thereof is connected to a lower core plate 74 having a disk shape and a plurality of flow holes (not illustrated) formed therein. Then, the lower core plate 74 is supported by the lower core support plate 70. That is, the core barrel 73 is supported in a suspended state on the lower core support plate 70 of the nuclear reactor vessel body 62. A reactor core 75 is formed by the upper core plate 72, the core barrel 73, and the lower core plate 74, and the reactor core 75 has a plurality of fuel assemblies 76 disposed therein. Although not illustrated in the drawings, each of the fuel assemblies 76 is formed by binding a plurality of fuel rods in a grid shape by a support grid. Here, the upper nozzle is fixed to the upper end, and the lower nozzle is fixed to the lower end. Further, the reactor core 75 has a plurality of control rods 77 disposed therein. The plurality of control rods 77 is formed as a control rod cluster 78 while the upper ends are evenly arranged, and is insertable into the fuel assembly 76. In the upper core support plate 69, a plurality of control rod cluster guide pipes 79 is fixed while penetrating the upper core support plate 69, and each control rod cluster guide pipe 79 is formed so that the lower end thereof extends to the control rod cluster 78 inside the fuel assembly 76. The upper portion of the nuclear reactor vessel head 63 that constitutes the nuclear reactor vessel 61 is formed in a semi-spherical shape, and a magnetic jack type control rod driving mechanism 80 is accommodated in a housing 81 which is integrated with the nuclear reactor vessel head 63. The plurality of control rod cluster guide pipes 79 is formed so that the upper ends thereof extend to the control rod driving mechanism 80, and control rod cluster driving shafts 82 which extend from the control rod driving mechanism 80 extend to the fuel assemblies 76 while passing through the inside of the control rod cluster guide pipes 79, thereby gripping the control rod cluster 78. The control rod driving mechanism 80 extends in the vertical direction so as to be connected to the control rod cluster 78, and a control rod cluster driving shaft 82 of which the surface is provided with a plurality of circumferential grooves formed equally pitched in the longitudinal direction is moved in the vertical direction by the magnetic jack, thereby controlling the output of the nuclear reactor. Further, the nuclear reactor vessel body 62 is provided with a plurality of instrumentation nozzles 83 which penetrates the lower end plate 66, and each of the instrumentation nozzles 83 is formed so that the upper end inside the reactor is connected to an in-core instrumentation guide pipe 84 and the lower end outside the reactor is connected to a conduit tube 85. In each of the in-core instrumentation guide pipes 84, the upper end is connected to the lower core support plate 70 and upper and lower connection plates 86 and 87 for suppressing a vibration are connected to the in-core instrumentation guide pipes. A thimble tube 88 is provided with a neutron flux detector (not illustrated) capable of measuring a neutron flux, and is insertable to the fuel assembly 76 while penetrating the lower core plate 74 from the conduit tube 85 along the instrumentation nozzle 83 and the in-core instrumentation guide pipe 84. Accordingly, the nuclear fission inside the reactor core 75 is controlled in a manner such that the control rod cluster driving shaft 82 is moved by the control rod driving mechanism 80 so as to extract the control rod 77 from the fuel assembly 76 by a predetermined amount. Then, the light water charged into the nuclear reactor vessel 61 is heated by the generated thermal energy, and the high-temperature light water is discharged from the outlet nozzle 68 so as to be sent to the steam generator 13 as described above. That is, neutrons are discharged by the nuclear fission of the atomic fuel forming the fuel assembly 76, and the light water as the moderator and the primary cooling water decreases the movement energy of the discharged high-speed neutrons so as to form thermal neutrons. Accordingly, new nuclear fission may easily occur, and the generated heat is removed and cooled. Meanwhile, when the control rod 77 is inserted into the fuel assembly 76, the number of neutrons generated inside the reactor core 75 may be adjusted. Further, when the entire control rod 77 is inserted into the fuel assembly 76, the nuclear reactor may be emergently stopped. Further, the nuclear reactor vessel 61 is formed so that an upper plenum 89 communicating with the outlet nozzle 68 is provided above the reactor core 75 and a lower plenum 90 is provided therebelow. Then, a down corner portion 91 which communicates with the inlet nozzle 67 and the lower plenum 90 is formed between the nuclear reactor vessel 61 and the core barrel 73. Accordingly, the light water flows from the inlet nozzle 67 into the nuclear reactor vessel body 62, flows downward to the down corner portion 91, reaches the lower plenum 90, rises while being guided upward by the spherical inner surface of the lower plenum 90, passes through the lower core support plate 70 and the lower core plate 74, and flows into the reactor core 75. The light water which flows into the reactor core 75 increases in temperature while cooling the fuel assembly 76 by absorbing the thermal energy generated from the fuel assembly 76 constituting the reactor core 75, passes through the upper core plate 72, rises to the upper plenum 89, and is discharged through the outlet nozzle 68. In the nuclear reactor vessel 61 with such a configuration, as illustrated in FIG. 13, the instrumentation nozzle 83 is formed in a manner such that an in-core instrumentation cylinder 95 is fitted into an attachment hole 96 formed in the lower end plate 66 of the nuclear reactor vessel body 62 and the upper end of the in-core instrumentation cylinder 95 is fixed to the inner surface (a groove-welding portion 97) of the lower end plate 66 by welding. The nuclear reactor vessel body 62 is formed by buttered-welding stainless steel to the inner surface of low-alloy steel, and the in-core instrumentation cylinder 95 of nickel alloy is welded to the nuclear reactor vessel body 62 by nickel alloy (as the groove-welding portion 97) while being fitted into the attachment hole 96 of the nuclear reactor vessel body 62. For that reason, there is a possibility that a tensile stress may remain in the in-core instrumentation cylinder 95, the groove-welding portion 97, and the vicinity thereof. Thus, there is an increase in the possibility of stress corrosion cracking due to the long-term use. Here, the stress corrosion cracking is prevented by solving the residual tensile stress of the surface using the residual compressive stress through a water jet peening (WJP) apparatus as a nuclear reactor repairing apparatus. The water jet peening apparatus is used to solve the residual tensile stress of the surface of the metal member by the residual compressive stress by jetting high-pressure water including cavitation air bubbles to the surface of the metal member under the water. Then, in a case where the residual tensile stress of the surface of the lower end plate 66 is solved by the residual compressive stress through the water jet peening apparatus, the water jet peening operation is performed while the water jet peening apparatus is attached to the instrumentation nozzle 83 (the in-core instrumentation cylinder 95). FIG. 1 is a longitudinal sectional view illustrating a clamping device of the water jet peening apparatus according to the first embodiment of the invention, FIG. 2 is a cross-sectional view taken along the line II-II of FIG. 1, FIG. 3 is a cross-sectional view taken along the line III-III of FIG. 1, FIG. 4 is a schematic diagram illustrating a water jet peening action for an inner surface, FIG. 5 is a schematic diagram illustrating a detection device of the water jet peening apparatus, FIG. 6 is a schematic diagram illustrating an action of the detection device, FIG. 7 is a schematic diagram illustrating a state where the water jet peening apparatus is provided, FIG. 8 is a schematic diagram illustrating a lower portion of a nuclear reactor vessel, FIG. 9 is a top view illustrating the lower portion of the nuclear reactor vessel, and FIG. 10 is a front view illustrating the water jet peening apparatus. In the first embodiment, as illustrated in FIG. 7, a water jet peening apparatus 101 is fixed to the instrumentation nozzle 83 (the in-core instrumentation cylinder 95) provided in the lower end plate (the semi-spherical portion) 66 of the nuclear reactor vessel 61 (the nuclear reactor vessel body 62). That is, in the nuclear power plant, a nuclear reactor building (not illustrated) is provided with a working floor 121, a cavity 122 is provided below the working floor 121, and cooling water is stored in the cavity 122. The nuclear reactor vessel 61 is disposed inside the cavity 122 while being supported in a suspended state. The nuclear reactor building is provided with an overhead traveling crane 123, and a hook 124 is movable and movable upward and downward in two directions intersecting the horizontal direction. Further, a pair of guide rails 125 is laid at both sides of the cavity 122 in the nuclear reactor building, and a movement crane 126 is supported so as to be movable. The movement crane 126 is provided with an electric hoist 127 which is movable in one direction (in FIG. 7, the left and right direction) of the horizontal direction and the other direction (in FIG. 7, the up and down direction perpendicular to the drawing paper) intersection (perpendicular to) one direction of the horizontal direction. Then, the electric hoist 127 includes a hook 128 which is movable upward and downward in the vertical direction. An installation pole 111 is an elongated member, and has a predetermined length. Here, the water jet peening apparatus 101 is connectable to the lower end thereof. The installation pole 111 includes a plurality of divided poles, and the divided poles may be coupled to each other while the flange portions are coupled to each other by a plurality of swing bolts in a close contact state. Further, in the embodiment, the installation pole 111 is used as the installation jig of the water jet peening apparatus 101, but the invention is not limited to this configuration. For example, a wire, a cable, or a rope may be used. Further, as illustrated in FIGS. 8 and 9, the nuclear reactor vessel 61 is supported by a concrete structure 131 provided on a firm ground such as a solid rock. The nuclear reactor vessel 61 is disposed in a cylindrical portion 132 provided in the concrete structure 131, and is supported in a suspended state. Then, the concrete structure 131 is located at the lower portion of the nuclear reactor vessel 61 and is provided with a thimble tube piping chamber 133. The plurality of conduit tubes 85 is curved so as to be drawn around the thimble tube piping chamber 133 while one ends are connected to the instrumentation nozzle 83 of the lower end plate 66 and the other ends are disposed in a monitoring chamber 134 in a collected state. A neutron flux detector 135 is attached to the front end of the thimble tube 88, is inserted from the other end of the conduit tube 85, and is disposed inside the nuclear reactor vessel 61 during the operation of the nuclear reactor. Then, the neutron flux detector 135 is extracted from the nuclear reactor vessel 61 through the thimble tube 88 during the water jet peening operation. As illustrated in FIG. 10, the water jet peening apparatus 101 includes an apparatus body 102, a clamping device 103, and an inner surface WJP nozzle 105. The clamping device 103 is disposed so as to protrude downward from the lower portion of the apparatus body 102, and is fitted and clamped to the outer peripheral surface of the instrumentation nozzle 83 (the in-core instrumentation cylinder 95) so that the apparatus body 102 is fixed to the instrumentation nozzle 83. The inner surface WJP nozzle 105 is used to jet high-pressure tensile water to the inner surface of the in-core instrumentation cylinder 95. In this case, the inner surface WJP nozzle 105 is movable upward and downward by an elevation mechanism (not illustrated) provided in the apparatus body 102, and is rotatable by a rotation mechanism (not illustrated). Accordingly, high-pressure water may be jetted to a predetermined area of the lower end plate 66, the groove-welding portion 97, and the in-core instrumentation cylinder 95. Further, the water jet peening apparatus 101 has a configuration in which the apparatus body 102 is provided with an operation monitoring camera 106 and apparatus positioning cameras 107 and 108. The operation monitoring camera 106 is fixed to the apparatus body 102, and is rotatable about the horizontal support shaft. Accordingly, the imaging direction may be changed. The apparatus positioning cameras 107 and 108 are separated from each other by a predetermined angle θ (for example, 90°) in the horizontal direction, and include illumination lamps. The apparatus positioning cameras 107 and 108 are movable upward and downward by an elevation cylinder 109 with respect to the apparatus body 102. Here, the apparatus positioning cameras are movable to the downward movement position during the positioning operation of the water jet peening apparatus 101, and are movable to the upward movement position during the water jet peening operation of the water jet peening apparatus 101. In the water jet peening apparatus 101, a connection shaft 110 is fixed to the upper portion of the apparatus body 102 and is connectable to the installation pole 111. In the clamping device 103 of the water jet peening apparatus 101, as illustrated in FIGS. 1 to 3, a clamping cylinder 201 has a cylindrical shape in which an upper clamping cylinder 202 and a lower clamping cylinder 203 are coupled to each other by a bolt 204, and is fixed to the apparatus body 102 by a bolt 205. A support cylinder 206 is disposed at the lower portion inside the clamping cylinder 201, and the lower end thereof is fixed to the lower end of the clamping cylinder 201 by a bolt 207. Further, a guide cylinder 208 is fitted into the support cylinder 206, and the lower end thereof is fixed to the lower end of the support cylinder 206 by a bolt 209. A plurality of (in the embodiment, three) clamping pieces 210 is disposed at the same interval in the circumferential direction of the support cylinder 206, and is supported by the support cylinder 206 and the guide cylinder 208 so as to be movable in the radial direction while being biased outward by a plate spring 211. A first piston guide 212 is fixed to the upper end of the support cylinder 206 by a bolt (not illustrated), a second piston guide 213 is fixed to the upper end of the first piston guide 212 by a bolt 214, and the upper end of the second piston guide 213 is fixed to the clamping cylinder 201 (the upper clamping cylinder 202) by a bolt 215. A first clamping piston 216 is disposed between the lower clamping cylinder 203 and the first piston guide 212, a second clamping piston 217 is disposed between the upper clamping cylinder 202 and the second piston guide 213, and the first clamping piston 216 and the second clamping piston 217 are integrally coupled to each other by a screw portion 218. The first clamping piston 216 and the second clamping piston 217 have a cylindrical shape, and the first clamping piston 216 has a configuration in which a slope surface 216a is formed in the inner peripheral portion so that the thickness of the lower end gradually decreases. Meanwhile, each clamping piece 210 is provided with a slope surface 210a which has a curved shape in which the inner surface side thereof follows the outer surface of the instrumentation nozzle 83 (the in-core instrumentation cylinder 95) and the outer surface side thereof is formed so that a gap with respect to the lower clamping cylinder 203 gradually increases upward. Then, the lower end of the first clamping piston 216 enters between each clamping piece 210 and the lower clamping cylinder 203, and the slope surface 216a contacts the slope surface 210a of each clamping piece 210. In the clamping cylinder 201, a chamber A is defined by the lower clamping cylinder 203, the first clamping piston 216, and the second clamping piston 217, and a first air supply port 219 for the chamber A is provided. Further, in the clamping cylinder 201, a chamber B is defined by the upper clamping cylinder 202, the lower clamping cylinder 203, and the second clamping piston 217, and a second air supply port 220 for the chamber B is formed. Accordingly, when air is supplied from the first air supply port 219 to the chamber A, the second clamping piston 217 moves downward along with the first clamping piston 216, and the slope surface 216a presses the slope surface 210a of each clamping piece 210. Accordingly, each clamping piece 210 moves inward so as to clamp the instrumentation nozzle 83 (the in-core instrumentation cylinder 95). Meanwhile, when air is supplied from the second air supply port 220 to the chamber B, the first clamping piston 216 moves upward through the second clamping piston 217, and the slope surface 216a is separated from the slope surface 210a of each clamping piece 210. Then, each clamping piece 210 moves outward by the biasing force of the plate spring 211 so as to release the clamping operation for the instrumentation nozzle 83 (the in-core instrumentation cylinder 95). In addition, the clamping mechanism of the invention capable of fixing the clamping cylinder 201 to the instrumentation nozzle 83 includes the first clamping piston 216, the second clamping piston 217, and each clamping piece 210. Further, the clamping cylinder of the invention includes the clamping cylinder 201, the support cylinder 206, the guide cylinder 208, and the piston guides 212 and 213. Further, the support cylinder 206 and the first piston guide 212 are provided so that a nozzle guide (a positioning member) 221 is supported so as to be movable upward and downward therein. The nozzle guide 221 has a cylindrical shape, and an inner surface WJP nozzle 105 is insertable thereinto. The nozzle guide 221 is provided with a guide surface 221a of which the diameter therein decreases, and hence may guide a connection rod 105a of the inner surface WJP nozzle 105. Further, the nozzle guide 221 is formed by integrating a cylindrical portion 222 and a flange portion 223 provided at the lower end of the cylindrical portion 222. Then, the nozzle guide 221 is provided with a plurality of drainage holes 224 provided at the same interval in the circumferential direction so as to radially penetrate the flange portion 223. Each drainage hole 224 is formed so that one end thereof communicates with the inside of the nozzle guide 221 and the other end thereof communicates with a space portion (a predetermined gap) S between the nozzle guide 221 and the support cylinder 206 (the guide cylinder 208). In addition, the space portion S is formed so that the lower portion thereof is opened to the outside. Accordingly, as illustrated in FIGS. 1 and 4, the water jet peening apparatus 101 moves downward, and the clamping device 103 moves downward to a predetermined position while being fitted to the outside of the instrumentation nozzle 83 (the in-core instrumentation cylinder 95). Then, the lower surface of the flange portion 223 of the nozzle guide 221 is stopped while contacting the upper end surface of the in-core instrumentation cylinder 95, so that the water jet peening apparatus 101 is positioned to a predetermined position of the instrumentation nozzle 83. Here, the instrumentation nozzle 83 (the in-core instrumentation cylinder 95) is clamped when each clamping piece 210 moves inward as described above. Further, a detection device 231 is provided which detects a state where the nozzle guide 221 is positioned to a predetermined position of the instrumentation nozzle 83 when the water jet peening apparatus 101 moves downward so as to be fixed to a predetermined position of the instrumentation nozzle 83 (the in-core instrumentation cylinder 95). In the detection device 231, as illustrated in FIGS. 5 and 6, the cylindrical nozzle guide 221 is supported so as to be movable upward and downward inside the support cylinder 206 and the first piston guide 212, and the lower end thereof is provide with the flange portion (the operation piece) 223. Further, a plurality of detection rods 232 is movably supported so as to penetrate the support cylinder 206 and the first piston guide 212 upward and downward at the outside of the nozzle guide 221, and the lower end thereof is able to contact the flange portion 223. A leading guide 233 is fixed to the lower portion of the second piston guide 213 (see FIG. 1), and the upper end of each detection rod 232 is fixed to a connection member 234 so as to penetrate the leading guide 233 upward. Then, a magnet 235 is fixed to the leading guide 233, and a magnet sensor 236 is fixed to the connection member 234. Accordingly, the water jet peening apparatus 101 moves downward, and moves downward to a predetermined position while the clamping device 103 is fitted to the outside of the instrumentation nozzle 83 (the in-core instrumentation cylinder 95). Then, the lower surface of the flange portion 223 of the nozzle guide 221 comes into surface-contact with the upper end surface of the in-core instrumentation cylinder 95, and moves upward while being pressed by the in-core instrumentation cylinder 95. Then, each detection rod 232 moves upward with respect to the downward moving leading guide 233, and the magnet sensor 236 of the connection member 234 moves to a position facing the magnet 235 of the leading guide 233. Here, when the magnet sensor 236 detects the magnet 235 and notifies the detection result to an operator, the downward movement of the water jet peening apparatus 101 is stopped, and the water jet peening apparatus 101 is positioned to a predetermined position of the instrumentation nozzle 83. Here, a method of performing a water jet peening operation on the inner surface of the instrumentation nozzle 83 (the in-core instrumentation cylinder 95) using the water jet peening apparatus 101 will be described. As illustrated in FIG. 7, the water jet peening apparatus 101 is suspended by the movement crane 126 through the installation pole 111 while cooling water is stored in the cavity 122. The water jet peening apparatus 101 moves therefrom in the horizontal direction through the movement crane 126, and the water jet peening apparatus 101 moves downward through the electric hoist 127 while being positioned to the instrumentation nozzle 83. Then, as illustrated in FIG. 6, the clamping cylinder 201 is fitted to the outside of the instrumentation nozzle 83, and the flange portion 223 of the nozzle guide 221 contacts the in-core instrumentation cylinder 95, so that the detection rod 232 moves upward along with the nozzle guide 221. Then, the detection device 231 detects this movement, and the operator stops the downward movement of the water jet peening apparatus 101. Here, when each clamping piece 210 moves inward, the instrumentation nozzle 83 is clamped, and the water jet peening apparatus 101 is fixed to the instrumentation nozzle 83. When the water jet peening apparatus 101 is fixed to the instrumentation nozzle 83, the inner surface WJP nozzle 105 moves to the inside of the in-core instrumentation cylinder 95 through the clamping device 103 as illustrated in FIG. 4. Then, the inner surface WJP nozzle 105 moves downward in a rotation state while high-pressure water is jetted therefrom so that the high-pressure water including cavitation air bubbles is jetted from the inner surface WJP nozzle 105 to the inner surface of the in-core instrumentation cylinder 95. Accordingly, the residual tensile stress of the inner surface of the in-core instrumentation cylinder 95 is solved by the residual compressive stress. Meanwhile, the high-pressure water which is jetted from the inner surface WJP nozzle 105 moves upward inside the instrumentation nozzle 83 since the neutron flux detector 135 or the thimble tube 88 exists inside the conduit tube 85. Then, the high-pressure water flows outward through each drainage hole 224, and moves downward in the space portion S between the nozzle guide 221 and the support cylinder 206 (the guide cylinder 208) so as to be discharged to the outside. In this way, the water jet peening apparatus of the first embodiment includes the clamping cylinder 201 which may be disposed at the outer peripheral side of the instrumentation nozzle 83 with a predetermined gap therebetween, the clamping piece 210 which may fix the clamping cylinder 201 to the instrumentation nozzle 83, the nozzle guide 221 which has a cylindrical shape, is provided inside the clamping cylinder 201, and is positioned to a position adjacent to the upper end of the instrumentation nozzle 83, the inner surface WJP nozzle 105 which may be movable upward and downward inside the nozzle guide 221, and the drainage hole 224 which radially penetrates the nozzle guide 221. Accordingly, when the high-pressure water including cavitation air bubbles is jetted from the inner surface WJP nozzle 105 to the inner surface of the instrumentation nozzle 83 under water, the residual tensile stress of the inner surface of the instrumentation nozzle 83 is solved by the residual compressive stress, and the jetted high-pressure water is discharged from the drainage hole 224 of the nozzle guide 221 while not substantially giving any action to the neutron flux detector 135. Accordingly, it is possible to improve the safety of the operation by preventing the thimble tube 88 from being popped out due to the water jet peening operation. In the water jet peening apparatus of the first embodiment, the lower end of the cylindrical portion 222 of the nozzle guide 221 is provided with the flange portion 223, and a plurality of drainage holes 224 is provided in the flange portion 223 at the same interval in the circumferential direction. Accordingly, the high-pressure water which is jetted from the inner surface WJP nozzle 105 is discharged highly efficiently from the plurality of drainage holes 224, and hence the draining performance may be improved. In the water jet peening apparatus of the first embodiment, the nozzle guide 221 may be positioned to a position contacting the upper end of the instrumentation nozzle 83. Accordingly, the nozzle guide 221 and the instrumentation nozzle 83 are disposed without any gap, and hence the high-pressure water which is jetted from the inner surface WJP nozzle 105 may be appropriately led to the plurality of drainage holes 224. The water jet peening apparatus of the first embodiment includes the detection device 231 which detects a state where the nozzle guide 221 is positioned to a predetermined position of the instrumentation nozzle 83. Accordingly, the water jet peening apparatus 101 may be highly precisely positioned to the instrumentation nozzle 83. In the water jet peening apparatus of the first embodiment, the detection rod 232 which is supported by the clamping cylinder 201 so as to be movable upward and downward and of which the lower end indirectly may contact the upper end of the instrumentation nozzle 83 through the flange portion 223 and the magnet sensor 236 which detects the up and down position of the detection rod 232 are provided as the detection device 231. Accordingly, when the flange portion 223 of the nozzle guide 221 contacts the upper end of the instrumentation nozzle 83, the detection rod 232 moves upward by the flange portion 223. Accordingly, the upward movement is detected by the magnet sensor 236, and hence the water jet peening apparatus 101 may be highly precisely positioned to the instrumentation nozzle 83 with a simple configuration. In the water jet peening apparatus of the first embodiment, the nozzle guide 221 is supported by the clamping cylinder 201 so as to be movable upward and downward, the flange portion 223 may come into surface-contact with the upper end surface of the instrumentation nozzle 83, and the detection rod 232 is movable upward and downward through the flange portion 223. Accordingly, when the flange portion 223 of the nozzle guide 221 comes into surface-contact with the upper end surface of the instrumentation nozzle 83, the detection rod 232 moves upward, and the movement of the nozzle guide 221 to a predetermined position is detected with high precision. Accordingly, the water jet peening apparatus 101 may be highly precisely positioned to the instrumentation nozzle 83. In the water jet peening apparatus of the first embodiment, the detection rod 232 is disposed outside the nozzle guide 221. Accordingly, since the inner surface WJP nozzle 105 moves upward and downward inside the nozzle guide 221 and the detection rod 232 is disposed outside the nozzle guide 221, the stable water jet peening operation may be performed while the detection rod 232 does not disturb the up-and-down movement of the inner surface WJP nozzle 105. FIG. 14 is a longitudinal sectional view illustrating a clamping device of a water jet peening apparatus according to a second embodiment of the invention, and FIG. 15 is a schematic diagram illustrating an action of a detection device. In addition, the same reference sign will be given to the member having the same function as the above-described embodiment, and the detailed description thereof will not be repeated. In the clamping device of the water jet peening apparatus according to the second embodiment, as illustrated in FIGS. 14 and 15, a clamping cylinder 301 includes a plurality of clamping pieces (clamping mechanisms) 302 therein and hence may clamp the instrumentation nozzle 83 (the in-core instrumentation cylinder 95) by the inward movement of a clamping piston (not illustrated). A nozzle guide (a positioning member) 303 is fixed into the clamping cylinder 301. The nozzle guide 303 has a cylindrical shape, and the inner surface WJP nozzle 105 is insertable thereinto. The nozzle guide 303 is obtained by integrating a cylindrical portion 304 and a flange portion 305 provided at the lower end of the cylindrical portion 304. Then, the nozzle guide 303 is provided with a plurality of drainage holes 306 provided at the same interval in the circumferential direction so as to radially penetrate the flange portion 305. Each drainage hole 306 has a configuration in which one end thereof communicates with the inside of the nozzle guide 303 and the other end thereof communicates with a space portion (a predetermined gap) S between the nozzle guide 303 and the clamping cylinder 301. In addition, the lower portion of the space portion S is opened to the outside. Accordingly, the water jet peening apparatus 101 (see FIG. 10) moves downward, and moves downward to a predetermined position while the clamping cylinder 301 is fitted to the outside of the instrumentation nozzle 83 (the in-core instrumentation cylinder 95). Then, the lower surface of the flange portion 305 of the nozzle guide 303 is stopped while contacting the upper end surface of the in-core instrumentation cylinder 95, and hence the water jet peening apparatus 101 is positioned to a predetermined position of the instrumentation nozzle 83. Here, the instrumentation nozzle 83 (the in-core instrumentation cylinder 95) may be clamped by the inward movement of the clamping piece 302. Further, a detection device 311 is provided which detects a state where the nozzle guide 303 is positioned to a predetermined position of the instrumentation nozzle 83 when the water jet peening apparatus 101 moves downward so as to be fixed to a predetermined position of the instrumentation nozzle 83 (the in-core instrumentation cylinder 95). In the detection device 311, a plurality of detection rods 312 is supported so as to be movable upward and downward while penetrating the flange portion 305 outside the nozzle guide 303, and each lower end may contact the upper end of the instrumentation nozzle 83 (the in-core instrumentation cylinder 95). A leading guide 313 is fixed to the clamping cylinder 301, and a connection member 314 is fixed to the upper end of each detection rod 312 while the detection rod penetrates the leading guide 313 upward. Then, a magnet sensor 315 is fixed to the leading guide 313, and a magnet 316 is fixed to the connection member 314. Accordingly, the water jet peening apparatus 101 moves downward, and moves downward to a predetermined position while the clamping cylinder 301 is fitted to the outside of the instrumentation nozzle 83 (the in-core instrumentation cylinder 95). Then, the lower end of the detection rod 312 contacts the upper end of the in-core instrumentation cylinder 95, and moves upward while being pressed by the in-core instrumentation cylinder 95. At this time, the lower surface of the flange portion 305 of the nozzle guide 303 comes into surface-contact with the upper end surface of the in-core instrumentation cylinder 95. Then, the detection rod 312 moves upward with respect to the downward moving clamping cylinder 301, and the magnet 316 of the connection member 314 moves to a position facing the magnet sensor 315 of the leading guide 313. Here, when the magnet sensor 315 detects the magnet 316 and notifies the detection result to the operator, the downward movement of the water jet peening apparatus 101 is stopped, and the water jet peening apparatus 101 is positioned to a predetermined position of the instrumentation nozzle 83. Here, a method of performing a water jet peening operation on the inner surface of the instrumentation nozzle 83 (the in-core instrumentation cylinder 95) using the water jet peening apparatus 101 will be described. When the water jet peening apparatus 101 moves downward, the clamping cylinder 301 is fitted to the outside of the instrumentation nozzle 83, and the lower end of the detection rod 312 moves upward while contacting the in-core instrumentation cylinder 95. Then, the detection device 311 detects this movement, and the operator stops the downward movement of the water jet peening apparatus 101. Here, the instrumentation nozzle 83 is clamped by the inward movement of each clamping piece 302, and hence the water jet peening apparatus 101 is fixed to the instrumentation nozzle 83. When the water jet peening apparatus 101 is fixed to the instrumentation nozzle 83, the inner surface WJP nozzle 105 moves downward into the in-core instrumentation cylinder 95. Then, the inner surface WJP nozzle 105 moves downward in a rotation state while high-pressure water is jetted therefrom so that the high-pressure water including cavitation air bubbles is jetted from the inner surface WJP nozzle 105 to the inner surface of the in-core instrumentation cylinder 95. Accordingly, the residual tensile stress of the inner surface of the in-core instrumentation cylinder 95 is solved by the residual compressive stress. Meanwhile, the jetted high-pressure water is discharged from the drainage hole 306 of the nozzle guide 303 to the space portion S while not substantially giving any action to the neutron flux detector 135, and moves downward so as to be discharged to the outside. In this way, the water jet peening apparatus of the second embodiment includes the clamping cylinder 301 which may be disposed at the outer peripheral side of the instrumentation nozzle 83 with a predetermined gap therebetween, the clamping piece 302 which may fix the clamping cylinder 301 to the instrumentation nozzle 83, the nozzle guide 303 which has a cylindrical shape, is provided inside the clamping cylinder 301, and is positioned to a position adjacent to the upper end of the instrumentation nozzle 83, the inner surface WJP nozzle 105 which is movable upward and downward inside the nozzle guide 303, and the drainage hole 306 which radially penetrates the nozzle guide 303. Accordingly, when the high-pressure water including cavitation air bubbles is jetted from the inner surface WJP nozzle 105 to the inner surface of the instrumentation nozzle 83 under water, the residual tensile stress of the inner surface of the instrumentation nozzle 83 is solved by the residual compressive stress, and the jetted high-pressure water is discharged from the drainage hole 306 of the nozzle guide 303 while not substantially giving any action to the neutron flux detector 135. Accordingly, it is possible to improve the safety of the operation by preventing the thimble tube 88 from being popped out due to the water jet peening operation. FIG. 16 is a schematic diagram illustrating an entire configuration of a water jet peening apparatus according to a third embodiment of the invention, FIG. 17 is a schematic diagram illustrating a fixing device, and FIG. 18 is a cross-sectional view taken along the line XVIII-XVIII of FIG. 17. In addition, the same reference sign will be given to the member having the same function as the above-described first embodiment, and the detailed description thereof will not be repeated. In the third embodiment, as illustrated in FIG. 16, the water jet peening apparatus 101 is fixed to the instrumentation nozzle 83 which is provided in the lower end plate (the semi-spherical portion) 66 of the nuclear reactor vessel 61 (the nuclear reactor vessel body 62). That is, the installation pole 111 extends downward from the working floor 121, and the water jet peening apparatus 101 is connected to the lower end thereof. The plurality of conduit tubes 85 is curved so as to be drawn around the thimble tube piping chamber 133 while one ends are connected to the instrumentation nozzle 83 of the lower end plate 66 and the other ends are disposed in a monitoring chamber 134 in a collected state. The conduit tube 85 is provided so that the other end extends to a seal table 351 provided in the monitoring chamber 134 so as to be connected thereto and the front end is provided with the thimble tube 88 to which a neutron flux detector (not illustrated) is attached. The thimble tube 88 is insertable from the monitoring chamber 134 into the other end of the conduit tube 85, and is insertable into the nuclear reactor vessel 61 through the instrumentation nozzle 83. Then, a fixing device 352 is provided which does not allow the end of the thimble tube 88, drawn from the other end of the conduit tube 85 in the monitoring chamber 134, to be movable with respect to the conduit tube 85. As illustrated in FIGS. 17 and 18, the fixing device 352 includes a first fixing jig 353 which is fixed to the end of the conduit tube 85, a second fixing jig 354 which is fixed to the thimble tube 88 drawn from the end of the conduit tube 85, and a connection member 355 which suppresses the separation of the first fixing jig 353 and the second fixing jig 354. That is, in the first fixing jig 353, the conduit tube 85 extends upward so that the end penetrates the seal table 351 and a nut 401 and a joint lower body 402 are fixed to the outer peripheral portion thereof. A metal fastener 403 has a U-shape in the lateral direction, and is fixed by a fixing nut 405 while a packing metal fastener 404 is threaded into an upper flange portion 403a, and a packing pressing metal fastener 406 is fixed to the lower portion thereof. The conduit tube 85 is nipped by a metal fastener 407 in a lower flange portion 403b of the metal fastener 403, and is fixed by a mounting screw 408. Then, a low-pressure sealing rubber packing 409 is fixed in a compressed state between the joint lower body 402 and the packing pressing metal fastener 406. In addition, the thimble tube 88 inside the conduit tube 85 is drawn upward from the packing metal fastener 404. In the second fixing jig 354, a rubber plate 411 is wound around the thimble tube 88, and the thimble tube is nipped by a first plate 412 having a flat shape and a second plate 413 having a shape of which the middle portion is curved. Then, the first plate 412 and the second plate 413 are integrally fixed to each other by a coupling bolt 414 and a coupling nut 415 provided at both sides of the nipped thimble tube 88. The connection member 355 is a soft rope such as a string, a cable, and a wire, and is wound around the first fixing jig 353 and the second fixing jig 354 so as to suppress the separation therebetween. In addition, the connection member 355 is not limited to the soft rope. For example, a hard connection rod or a connection plate may be used. Further, a monitoring device 361 which monitors the fixed state of the thimble tube 88 and the conduit tube 85 using the fixing device 352 is provided. The monitoring device 361 includes a camera 362 which is disposed inside the monitoring chamber 134 so as to capture the image of the fixing device 352 and a monitoring unit 364 which is provided in a working chamber 363 provided in the working floor 121. Then, the monitoring unit 364 is connected to an operation device 365 and a display (a display device) 366. Then, the camera 362 may output a captured image to the monitoring unit 364, and the operator in the working chamber 363 may monitor the fixed state of the thimble tube 88 and the conduit tube 85 using the fixing device 352 through the display 366. Here, a method of performing a water jet peening operation on the inner surface of the instrumentation nozzle 83 using the water jet peening apparatus 101 will be described. As illustrated in FIGS. 1, 4, and 16, the water jet peening method of the embodiment includes: disposing the clamping cylinder 201 at the outer peripheral side of the instrumentation nozzle 83 with a predetermined gap therebetween, fixing the nozzle guide (the positioning member) 221 provided in the clamping cylinder 201 to the instrumentation nozzle 83 at a position adjacent to the upper end of the instrumentation nozzle 83, allowing the thimble tube 88 drawn from the instrumentation nozzle 83 to the outside through the conduit tube 85 to be immovable, moving the inner surface WJP nozzle 105 downward to the instrumentation nozzle 83 through the clamping cylinder 201, jetting high-pressure water to the inner surface of the instrumentation nozzle 83 by moving the inner surface WJP nozzle 105 downward in a rotation state while jetting the high-pressure water including cavitation air bubbles from the inner surface WJP nozzle 105, and discharging the high-pressure water which is jetted from the inner surface WJP nozzle 105 from the drainage hole 224 provided in the nozzle guide 221 to the outside. Further, the fixed state of the thimble tube 88 and the conduit tube 85 is monitored when the high-pressure water is jetted from the inner surface WJP nozzle 105 to the inner surface of the instrumentation nozzle 83. In this way, in the water jet peening apparatus of the third embodiment, the fixing device 352 is provided which does not allow the end of the thimble tube 88 drawn from the other end of the conduit tube 85 to be movable with respect to the conduit tube 85. Accordingly, since the thimble tube 88 which is drawn from the other end of the conduit tube 85 is immovably fixed by the fixing device 352, it is possible to reliably prevent the thimble tube 88 from being popped out due to the water jet peening operation. In the water jet peening apparatus of the third embodiment, the first fixing jig 353 which is fixed to the conduit tube 85 extending inside the monitoring chamber 134, the second fixing jig 354 which is fixed to the end of the thimble tube 88 drawn from the other end of the conduit tube 85, and the connection member 355 which suppresses the separation between the first fixing jig 353 and the second fixing jig 354 are provided as the fixing device 352. Accordingly, it is possible to easily prevent the thimble tube 88 from being popped out due to the water jet peening operation with a simple configuration. In the water jet peening apparatus of the third embodiment, the monitoring device 361 is provided which monitors the fixed state of the thimble tube 88 and the conduit tube 85. Accordingly, since the fixed state of the thimble tube 88 and the conduit tube 85 is monitored by the monitoring device 361, it is possible to prevent the thimble tube 88 from being popped out due to the water jet peening operation. Further, the water jet peening method of the third embodiment includes: disposing the clamping cylinder 201 at the outer peripheral side of the instrumentation nozzle 83 with a predetermined gap therebetween, fixing the nozzle guide (the positioning member) 221 provided in the clamping cylinder 201 to the instrumentation nozzle 83 at a position adjacent to the upper end of the instrumentation nozzle 83, allowing the thimble tube 88 drawn from the instrumentation nozzle 83 to the outside through the conduit tube 85 to be immovable, moving the inner surface WJP nozzle 105 downward to the instrumentation nozzle 83 through the clamping cylinder 201, jetting high-pressure water to the inner surface of the instrumentation nozzle 83 by moving the inner surface WJP nozzle 105 downward in a rotation state while jetting the high-pressure water including cavitation air bubbles from the inner surface WJP nozzle 105, and discharging the high-pressure water which is jetted from the inner surface WJP nozzle 105 from the drainage hole 224 provided in the nozzle guide 221 to the outside. Accordingly, when the high-pressure water including cavitation air bubbles is jetted from the inner surface WJP nozzle 105 to the inner surface of the instrumentation nozzle 83 under water, the residual tensile stress of the inner surface of the instrumentation nozzle 83 is solved by the residual compressive stress, and the jetted high-pressure water is discharged from the drainage hole 224 of the nozzle guide 221 while not substantially giving any action to the neutron flux detector 135. Accordingly, it is possible to improve the safety of the operation by preventing the thimble tube 88 from being popped out due to the water jet peening operation. Further, since the high-pressure water is jetted while the thimble tube 88 drawn to the outside from the instrumentation nozzle 83 through the conduit tube 85 is immovably fixed, it is possible to reliably prevent the thimble tube 88 from being popped out due to the water jet peening operation. In the water jet peening method of the third embodiment, the fixed state of the thimble tube 88 and the conduit tube 85 is monitored when the high-pressure water is jetted from the inner surface WJP nozzle 105 to the inner surface of the instrumentation nozzle 83. Accordingly, it is possible to prevent the thimble tube 88 from being popped out due to the water jet peening operation. 61 NUCLEAR REACTOR VESSEL 62 NUCLEAR REACTOR VESSEL BODY 63 NUCLEAR REACTOR VESSEL HEAD 66 LOWER END PLATE (SEMI-SPHERICAL PORTION) 83 INSTRUMENTATION NOZZLE 85 CONDUIT TUBE 88 THIMBLE TUBE 95 IN-CORE INSTRUMENTATION CYLINDER 101 WATER JET PEENING APPARATUS (NUCLEAR REACTOR REPAIRING APPARATUS) 102 APPARATUS BODY 103 CLAMPING DEVICE 105 INNER SURFACE WJP NOZZLE 201, 301 CLAMPING CYLINDER 206 SUPPORT CYLINDER 208 GUIDE CYLINDER 210, 302 CLAMPING PIECE (CLAMPING MECHANISM) 212 FIRST PISTON GUIDE 213 SECOND PISTON GUIDE 216 FIRST CLAMPING PISTON 217 SECOND CLAMPING PISTON 221, 303 NOZZLE GUIDE (POSITIONING MEMBER) 231, 311 DETECTION DEVICE 352 FIXING DEVICE 353 FIRST FIXING JIG 354 SECOND FIXING JIG 355 CONNECTION MEMBER 361 MONITORING DEVICE |
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description | This application is based upon and claims the benefit of priority from Japanese patent application No. 2007-103094, filed on Apr. 10, 2007, the disclosure of which is incorporated herein in its entirety by reference. The present invention relates to an ion implantation apparatus including an injector flag Faraday cup having a function of shutting off an ion beam as necessary and measuring a beam current. As a method of forming a conductive layer of an n type or a p type at a semiconductor wafer, there is used a so-called ion implantation technology of accelerating a conductive type dopant ionized by an ion source by an electric field to implant to a wafer. According to the ion implantation technology, a plasma is generated by ionizing a gas supplied to an ion source in a plasma chamber, and an ion beam is extracted from the plasma by applying a predetermined voltage to an extracting electrode. Successively, an ion beam comprising an ion having a desired mass is extracted by making the extracted ion beam incident on a mass analysis magnet apparatus, the ion beam is made to pass through a mass analysis slit, thereafter, reciprocally scanned by a beam scanner and irradiated to a wafer to thereby carry out ion implantation (see Patent Reference 1: JP-A-2006-156259). In such an ion implantation technology, there is provided an injector flag Faraday cup having a function of shutting off an ion beam as necessary and measuring a total beam current to be able to be brought in and out to and from a beam line. Graphite is provided at a portion of the injector flag Faraday cup on which an ion beam impinges, and when the injector flag Faraday cup is inserted into the beam line, the ion beam is shut off by making the ion beam impinge on the graphite. Specifically, as shown by FIG. 1A and FIG. 1B, an injector flag Faraday cup 200 is provided at inside of a scanner housing 310 along with a beam scanner 300. As explained later, the beam scanner 300 is for periodically reciprocating an incident ion beam in a horizontal direction orthogonal to an advancing direction thereof by a pair of scanning electrodes 300-1 and 300-2 arranged to be opposed to each other to interpose a beam trajectory line. An upstream side and a downstream side of the beam scanner 300 are respectively provided with scanner suppression electrodes 320 and 330 for restraining diversion of the ion beam and restricting a section size of the ion beam. The injector flag Faraday cup 200 is arranged at a portion in correspondence with a downstream side adjacent to the scanner suppression electrodes 330. The injector flag Faraday cup 200 is provided with a receiving area in correspondence with a range of scanning the ion beam by the beam scanner 300 and is made to be brought in and out to and from the beam line by a drive mechanism (not illustrated) installed at outside of the scanner housing 310, in this case, by being driven in an up and down direction. For example, during a time period until interchanging a wafer finished with ion implantation by a wafer which has not been implanted yet, the injector flag Faraday cup 200 is placed at the beam line to shut off the ion beam. A portion of the injector flag Faraday cup 200 on which the ion beam impinges is covered by a material of graphite or the like which is strong at sputtering by the ion beam. However, when the ion beam impinges on a graphite inner wall member at the injector flag Faraday cup 200, there is a case of bringing about sputtering of graphite. There is a case in which a sputtered graphite particle is adhered to peripheral members on an upstream side of the injector flag Faraday cup 200, particularly to the scanner suppression electrodes 330 or the scanning electrodes 300-1 and 300-2 to contaminate, further, a part of the scanner suppression electrodes 330 and the scanning electrodes 300-1 and 300-2 are exfoliated by secondary sputtering. When the scanner suppression electrodes 330 or the scanning electrodes 300-1 and 300-2 are contaminated or exfoliated in this way, there is a concern that the ion beam cannot reciprocally be scanned accurately. Further, when a large amount of graphite is adhered to between the scanner housing 310 and the scanner suppression electrodes 330 to bring about short-circuit, the ion beam cannot reciprocally be scanned. The present invention has been carried out in view of such a problem and it is an object thereof to provide an ion implantation apparatus in which a peripheral member thereof is not effected with an adverse influence by an injector flag Faraday cup and which can particularly maintain the beam scanner in an optimum state. An ion implantation apparatus according to the present invention includes a beam line for implanting an ion to a wafer by irradiating an ion beam extracted from an ion source and passed through a mass analysis magnet apparatus and a mass analysis slit to the wafer by being reciprocally scanned by a beam scanner. According to an aspect of the present invention, the beam line after passing the mass analysis slit before incidence of the beam scanner is arranged with a Faraday cup that detects a beam current by measuring a total beam amount of the ion beam to be able to be brought in and out thereto and therefrom. In the ion implantation apparatus according to the present invention, it is preferable that further includes a scanner housing that contains the beam scanner and the Faraday cup. In this case, the Faraday cup is arranged immediately after an ion beam inlet at the scanner housing and the beam scanner is arranged immediately after the Faraday cup. In the ion implantation apparatus according to the present invention, it is preferable that a shape of a beam incident portion at the Faraday cup is constituted by a rectangular shape to be able to deal with an ion beam having a section in an elliptical shape having a long axis in a lateral direction or a longitudinal direction. In the ion implantation apparatus according to the present invention, it is preferable that a drive mechanism which brings in and out the Faraday cup to and from the beam line is installed at outside of the scanner housing and the Faraday cup is attached to a drive shaft of the drive mechanism introduced into the scanner housing by penetrating an wall of the scanner housing. In the ion implantation apparatus according to the present invention, it may further includes a beam dump arranged at a most downstream position of the beam line and having a beam current detecting function. In this case, a beam transporting efficiency is made to be able to be calculated by comparing a detected value of the Faraday cup and a detected value of the beam dump. In the ion implantation apparatus according to the present invention, it may further include a profile monitor which measuring a current density distribution of a section of the ion beam. In this case, the profile monitor is arranged at a immediate vicinity on an upstream side or an immediate vicinity on a downstream side of the Faraday cup at inside of the scanner housing. In the ion implantation apparatus according to the present invention, it may further include a dose amount measurement unit arranged at a vicinity of the wafer, a determination portion that determines whether the measured dose amount is proper, a deflecting apparatus arranged at a section of the beam line from an outlet of the mass analysis magnet apparatus to a front side of the mass analysis slit for deflecting the ion beam in a predetermined direction deviated from the beam trajectory line and maintaining the deflection, and a control portion that carries out temporal evacuating by the deflecting apparatus when the dose amount measured in implanting the ion is determined to be improper by the determination portion. In this case, the control portion recovers the ion beam to the beam trajectory line by stopping the temporal evacuating when a predetermined time period has elapsed since the dose amount has been determined to be improper, and when the dose amount remeasured by the dose amount measurement portion is determined to be improper again, the control portion inserts the Faraday cup to the beam line and releases the temporal evacuating. In the ion implantation apparatus according to the present invention, the injector flag Faraday cup is inserted to the beam line on the upstream side of the beam scanner. Thereby, when the ion beam extracted from the ion source is shut off by the injector flag Faraday cup, the ion beam impinges on the injector flag Faraday cup. At this occasion, the ion beam is shut off by the injector flag Faraday cup arranged on the upstream side of the beam scanner, and therefore, even when sputtering is brought about by the ion beam, sputtered particles do not adhere to the peripheral member, for example, the scanning electrodes of the beam scanner. Therefore, the scanning electrodes of the beam scanner can be maintained in an optimum state, as a result, the ion beam can accurately be scanned reciprocally by the beam scanner. In addition thereto, a large amount of sputtered particles do not adhere to the scanning electrodes of the beam scanner, and therefore, a concern of short-circuiting the scanner housing containing the beam scanner and the like and the scanning electrodes of the beam scanner can firmly be prevented. Further, the injector flag Faraday cup can be made to be smaller than that of a constitution of arranging the injector flag Faraday cup at a vicinity of a downstream side of the beam scanner. This is because according to the constitution arranging the injector flag Faraday cup on the downstream side of the beam scanner, there is needed an injector flag Faraday cup having a receiving area adapted to a range of scanning the ion beam by the beam scanner. In contrast thereto, because it is unnecessary to prepare an injector flag Faraday cup having a receiving area adapted to the range of scanning the ion beam since according to the constitution of arranging the injector flag Faraday cup on the upstream side of the beam scanner, the injector flag Faraday cup is placed before the place being reciprocally scanned with the ion beam by the beam scanner. As a result, by only constructing a constitution of arranging the injector flag Faraday cup on the upstream side of the beam scanner, the constitution can contribute to small-sized formation of the injector flag Faraday cup. An embodiment of an ion implantation apparatus according to the present invention will be explained in reference to the drawings as follows. FIG. 2A and FIG. 2B are schematic views when the present invention is applied to an ion implantation apparatus of a single wafer type, particularly, FIG. 2A is a plane view and FIG. 2B is a side view. A constitution of the ion implantation apparatus 1 will be explained from the most upstream side of a beam line constituting a start point by an ion source 10. An outlet side of the ion source 10 is provided with an extracting electrode 12 for extracting an ion beam from a plasma generated at inside of an ion chamber. A vicinity of a downstream side of the extracting electrode 12 is provided with a suppression electrode 14 for restraining an electron included in the ion beam extracted from the extracting electrode 12 from flowing back to the extracting electrode 12. The ion source 10 is connected with an ion source high voltage power source 16 and an extracting power source 20 is connected to between the extracting electrode 12 and a terminal 18. A downstream side of the extracting electrode 12 is arranged with a mass analysis magnet apparatus 22 for extracting an ion beam comprising a desired ion by separating an ion other than the desired ion from an incident ion beam. A downstream side of the mass analysis magnet apparatus 22 is arranged with a first quadrupole vertically focusing electromagnet 24 for focusing or converging an ion beam in a longitudinal (vertical) direction, a park electrode (deflecting apparatus) 26 for deflecting an ion beam from a beam trajectory lines a mass analysis slit 28 for passing an ion beam comprising an ion of a desired mass in the ion beam, and a second quadrupole vertically focusing electromagnet 30 for focusing or converging an ion beam in a longitudinal direction. The park electrode 26 and the mass analysis slit 28 are contained in a park housing 27 constituted by a material in which cross contamination of aluminum or the like is hardly present. Further, as the mass analysis slit 28, other than an exclusive slit of a fixed type, a plurality of stages of switching type mass analysis slits may be used. According to the plurality of stages of the switching type mass analysis slits, three stages of slit sizes of, for example, an elliptical/or an oval type slit for high beam current, a long and narrow circular slit for a low beam current, and an extremely small diameter slit for confirming a beam trajectory axis are mechanically switched. A downstream side of the second quadrupole vertically focusing electromagnet 30 is arranged with an injector flag Faraday cup 32 for shutting off an ion beam as necessary and measuring a beam current, and a beam scanner 36 for periodically reciprocating to scan the ion beam in a horizontal direction orthogonal to a direction of advancing the ion beam. An upstream side and a downstream side of the beam scanner 36 are respectively provided with scanner suppression electrodes 34 and 38 having openings capable of restricting also a size of a sectional size of the ion beam, restraining diversion of the ion beam and shielding a scanning electric field from a surrounding. Further, the injector flag Faraday cup 32 is made to be able to be inserted and taken out to and from the beam line by a drive mechanism in an up and down direction in this case as explained later. Further, the injector flag Faraday cup 32, the beam scanner 36 and the scanner suppression electrodes 34 and 38 are contained in a scanner housing 37 made of aluminum. Respective members on the beam line from the extracting electrode 12 to the scanner housing 37 are contained in the terminal 18. The terminal 18 is connected with a terminal power source 19. Therefore, potentials of the park housing 27 and the scanner housing 37 are the same as a potential of the terminal 18 to constitute the potential of the terminal power source 19. A downstream side of the beam scanner 36 is arranged with a parallel lens 40 for redeflecting an ion beam deflected to have an angle in a horizontal direction relative to a center trajectory (center trajectory of ion beam before being scanned by the beam scanner 36) to be in parallel with the center trajectory, and an accelerating/decelerating column 42 for accelerating or decelerating the ion beam. The parallel lens 40 is constituted by a plurality of electrodes in a circular arc shape bored with holes for passing the ion beam at centers thereof. A first electrode from an upstream side of the parallel lens 40 is maintained at the terminal potential. A second electrode is referred to as a suppression electrode for restraining an electron from flowing in by being connected with the suppression power source 44. A third electrode is connected with a parallel lens power source 46, thereby, an electric field is generated between the second electrode and the third electrode, and an ion beam deflected in a horizontal direction becomes an ion beam in parallel with a center trajectory before being deflected. The parallel lens 40 is constructed by a structure of utilizing the electric field and the ion beam is decelerated by a potential difference between the second electrode and the third electrode. That is, the ion beam deflected by the beam scanner 36 is corrected in a trajectory thereof in a direction in parallel with a center trajectory before being deflected by the electric field between the second electrode and the third electrode and decelerated. The accelerating/decelerating column 42 is constituted by one or more of electrodes in a linear shape. A first electrode from an upstream side of the accelerating/decelerating column 42 is connected with the parallel lens power source 46 similar to the third electrode of the parallel lens 40. Second and third electrodes are respectively connected with a first accelerating/decelerating column power source 48 and a second accelerating/decelerating column power source 50. The ion beam is accelerated or decelerated by adjusting voltages of the power sources. Further, a fourth electrode is grounded to a ground potential. A downstream side of the accelerating/decelerating column 42 is arranged with an angular energy filter (hereinafter, referred to as AEF) 52 of a hybrid type. AEF 52 is an energy filter for selecting the ion beam achieving an aimed acceleration energy. AEF 52 includes a magnetic deflecting electromagnet for magnetic field deflection and a static deflecting electrode for static deflection. The magnetic deflecting electromagnet is arranged to surround an AEF chamber 54 and is constituted by a yoke member surrounding upper and lower and left and right sides of the AEF chamber 54 and a group of coils wound around the yoke member. Further, the magnetic deflecting electromagnet is connected with a direct current voltage power source (not illustrated). On the other hand, the static deflecting electrode is constituted by a pair of upper and lower AEF electrodes 56 and arranged to interpose an ion beam from up and down directions. In the pair of AEF electrodes 56, the AEF electrode 56 on an upper side is applied with a positive voltage and the AEF electrode 56 on a lower side is applied with a negative voltage, respectively. In deflecting by a magnetic field, an ion beam is deflected to a lower side by about 20 degrees by the magnetic field from the magnetic deflecting electromagnet and only an ion beam of an aimed energy is selected. On the other hand, in deflecting by the magnetic field and the electric field, or only the electric field, the ion beam is deflected to the lower side by about 20 degrees by a combining operation by the magnetic field from the magnetic deflecting electromagnet and the electric field generated between the pair of AEF electrodes 56, or a deflecting operation of the electric field and only an ion beam of an aimed energy is selected. In this way, AEF 52 is of the hybrid type using the magnetic field, the electric field, and both of the magnetic field and the electric field as necessary, and therefore, in transporting a low energy beam, the magnetic field having a high electron confining effect can mainly be used and in transporting a high energy beam, in addition to using both of the magnetic field deflection and the static deflection, a deflecting operation of only the electric field can be used. Further, a way of use differs by an energy or a kind of a gas of the ion source 10 when the magnetic field is always used, or when both of the magnetic field and the electric field is used or the deflecting operation of only the electric field is used. AEF 52 is provided with an AEF plasma shower 60 for promoting an efficiency of transporting an ion beam to a wafer 58 by restraining diversion of the ion beam by supplying an electron. Further, AEF 52 is respectively provided with AEF suppression electrodes 62 and 64 on an upstream side and a downstream side of the AEF plasma shower 60. The AEF suppression electrodes 62 and 64 mainly serve to restrict an electron barrier and a size of a sectional shape of the ion beam. A wall of the AEF chamber 54 is arranged with a plurality of permanent magnets 66 for forming a cusp magnetic field. By forming the cusp magnetic field, an electron is confined to inside of the AEF chamber 54. The respective permanent magnets 66 are arranged such that magnetic poles thereof are directed to inside of the AEF chamber 54 and the contiguous magnetic poles have opposite magnetic poles. Further, an outlet side of the AEF chamber 54 is provided with a striker plate 68 for receiving a neutral particle or the like constituted by neutralizing an ion advancing straight without being deflected by AEF 52. A processing chamber (vacuum processing chamber) 70 is connected with the AEF chamber 54. Selectable energy slits (hereinafter, referred to as SES) 72 are arranged at inside of the processing chamber 70. The selectable energy slits 72 are arranged to interpose the ion beam from up and down directions. Upper and lower selectable slits each includes 4 of slit faces, after selecting the slit face, by further adjusting axes of the upper and lower selectable slits in the up and down direction, and rotating the axes, a desired slit width is provided. By successively selecting 4 of the silt faces in accordance with a species of an ion, cross contamination is reduced. A plasma shower 74 supplies a low energy electron to a front face of the wafer 58 along with the ion beam, neutralizes and restrains charge up of a positive charge produced by ion implanting. Dose cups 76 respectively arranged at left and right ends of the plasma shower 74 measure a dose amount. Specifically, the dose cup is connected with a current measurement circuit and the dose amount is measured by measuring a beam current which flows by making the ion beam incident thereon. A beam profiler 78 includes a beam profiler cup (not illustrated) for measuring the beam current at an ion implanting position and a vertical profile cup (not illustrated) for measuring a beam shape and a beam X-Y position. The beam profiler 78 measures an ion beam density at the ion implanting position while being moved in a horizontal direction before implanting an ion or the like. When a predicted non uniformity (PNU) of the ion beam does not satisfy a request of the process as a result of measuring the beam profile, an applied voltage or the like of the beam scanner 36 is automatically adjusted to satisfy a process condition. The vertical profiler cup confirms a beam width and a beam center position by measuring the beam shape at the ion implanting position. The most downstream side of the beam line is arranged with a triple surface beam dump (TSBD) 80 having a beam current measurement function similar to that of a Faraday cup for measuring a final setup beam. The triple surface beam dump 80 reduces cross contamination by switching three faces of a triangular pillar in accordance with a kind of a gas of the ion source 10. Further, the beam line is naturally maintained in high vacuum. An explanation will be given of the injector flag Faraday cup 32 of the ion implantation apparatus 1 constituted as described above in reference to FIG. 3A and FIG. 3B. FIG. 3A is a side sectional view showing the injector flag Faraday cup 32 and a structure of a periphery thereof, and FIG. 3B is a plane sectional view thereof. In the ion implantation apparatus 1 according to the embodiment, the injector flag Faraday cup 32 is arranged on an upstream side of the beam scanner 36 and at inside of the scanner housing 37. Specifically, the injector flag Faraday cup 32 is arranged immediately after an ion beam inlet at the scanner housing 37 and the beam scanner 36 is arranged immediately after the injector flag Faraday cup 32. As shown by FIG. 3B, the beam scanner 36 is provided with a pair of scanning electrodes 36a and 36b arranged to increase a distance of being separated from each other along a beam trajectory line indicated by a one-dotted chain line. Here, a drive mechanism 32-1 is provided on an outer side of the scanner housing 37 to drive the injector flag Faraday cup 32 in an up and down direction to be brought in and out to and from the beam line. A drive shaft 32-2 is introduced from the drive mechanism 32-1 into the scanner housing 37 and the injector flag Faraday cup 32 is attached to a lower end of the drive shaft 32-2. In order to prevent a vacuum state (reduced pressure state) at inside of the scanner housing 37 from being deteriorated, the drive mechanism 32-1 is contained at inside of a housing 32-3 to be maintained in an airtight state and also a surrounding of the drive shaft 32-2 penetrating the scanner housing 37 is sealed. The injector flag Faraday cup 32 is used for measuring a current of the ion beam, normally arranged at an evacuating position indicated by a bold line in FIG. 3A, and is placed on the beam trajectory line as indicated by a two-dotted chain line in the drawing by being moved down in measuring. A principle of measuring the beam current by the injector flag Faraday cup 32 is as follows. The injector flag Faraday cup 32 is grounded by way of a terminal beam monitor controller (not illustrated). When the injector flag Faraday cup 32 is placed on the beam trajectory line, an electron in correspondence with the ion advancing to the injector flag Faraday cup 32 flows from the ground to the injector flag Faraday cup 32 to neutralize the ion. The terminal beam monitor controller measures an amount of electrons flowing for neutralizing the ion and calculates a beam current amount. The injector flag Faraday cup 32 may be provided with a receiving area in correspondence with a sectional shape of an ion beam incident thereon. That is, in the case of the example, although since the sectional shape of the ion beam incident on the injector flag Faraday cup 32 is constituted by an elliptical or a flat shape having a long axis in a lateral direction, and the injector flag Faraday cup is located before the location where the ion beam is reciprocally scanned, a beam receiving area of the injector flag Faraday cup 32 may be constituted by a rectangular shape slightly larger than a sectional shape of the ion beam. Graphite 32a is provided at a portion of the injector flag Faraday cup 32 on which the ion beam impinges, specifically, a face utilized for detecting the beam current. When the injector flag Faraday cup 32 is moved down to be arranged on the beam trajectory line, the ion beam advances to the injector flag Faraday cup 32 to impinge on the graphite 32a. At this occasion, even when the graphite 32a is sputtered by the ion beam, since the injector flag Faraday cup 32 is arranged on the upstream side of the beam scanner 36 and the ion beam is shut off by the injector flag Faraday cup 32, particles of the scattered graphite 32a are not adhered to the scanning electrodes 36a and 36b of the beam scanner 36. Therefore, the scanning electrodes 36a and 36b of the beam scanner 36 can be maintained in an optimum state, as a result, the ion beam can accurately be scanned reciprocally by the beam scanner 36. Further, a large amount of particles of the graphite 32a do not adhere to the scanning electrodes 36a and 36b of the beam scanner 36, and therefore, the concern of short-circuiting the scanner housing 37 containing the beam scanner 36 and the like and the scanning electrodes 36a and 36b of the beam scanner 36 can firmly be prevented. Further, the injector flag Faraday cup 32 can be made to be smaller than that of the related art constitution of arranging the injector flag Faraday cup 200 on the downstream side of the beam scanner 300. This is due to the following reason. According to the related art constitution of arranging the injector flag Faraday cup on the downstream side of the beam scanner, there is needed an injector flag Faraday cup having a receiving area adapted to a wide range of reciprocally scanning the ion beam. In contrast thereto, according to the constitution of the embodiment of arranging the injector flag Faraday cup 32 on the upstream side of the beam scanner 36, the ion beam is not reciprocally scanned, and therefore, the receiving area of the injector flag Faraday cup 32 can be made to be small. Particularly, the ion beam incident on the injector flag Faraday cup 32 is converged by the quadrupole vertically focusing electromagnet 30 arranged at a prestage thereof, and therefore, the sectional shape is further reduced. As a result thereof, a size of a total of the injector flag Faraday cup 32 can be reduced. On the other hand, when the injector flag Faraday cup 32 is moved up to evacuate from the ion beam, the ion beam advances to the beam scanner 36 arranged on the downstream side. The scanning electrodes 36a and 36b of the beam scanner 36 are maintained in the optimum state, and therefore, the ion beam arriving at the beam scanner 36 is accurately scanned reciprocally. Further, a beam transporting efficiency can be calculated by comparing a detected value of the injector flag Faraday cup 32 and a detected value of a triple surface beam dump 80. FIG. 4 is a plane sectional view showing a behavior of reciprocal scanning of the ion beam at inside of the scanner housing 37. Next, an explanation will be given of a function of shutting off the ion beam carried out by combining the park electrode 26 and the injector flag Faraday cup 32. FIG. 5 is a vertical sectional view showing the park electrode 26 and the mass analysis slit 28 and the park housing 27 containing these. In FIG. 5, at the beam trajectory line indicated by a two-dotted chain line, the park electrode 26 comprising a plus electrode 26-1 and a minus electrode 26-2 is arranged on a front side of the mass analysis slit 28, that is, on the upstream side. The park electrode 26 and the mass analysis slit 28 are contained in the park housing 27 comprising aluminum. A center of the mass analysis slit 28 is provided with a hole 120 for passing an ion beam comprising an ion of a predetermined mass in the ion beam. A face on the upstream side of the mass analysis slit 28, a wall face of the hole 120 and an inner wall face of the park housing 27 in correspondence with a downstream side of the minus electrode 26-2 are covered by a graphite 122. The graphite 122 is difficult to be sputtered and difficult to be exfoliated even when the ion beam impinges thereon. In FIG. 5, when a park voltage is not applied to the park electrode 26 and there is not a potential difference between the plus electrode 26-1 and the minus electrode 26-2 and the electric field is not present at the park electrode 26, the ion beam from the mass analysis magnet apparatus 22 passes the park electrode 26 along the beam trajectory line. Among the ion beam having passed the park electrode 26, the ion beam composed of ions of a predetermined mass passes the hole 120 of the mass analysis slit 28. The ion beam passing the hole 120 of the mass analysis slit 28 advances to the beam scanner 36 disposed on the downstream side (refer to FIG. 2A). On the other hand, when the electric field is present at the park electrode 26 comprising the plus electrode 26-1 and the minus electrode 26-2 by applying the park voltage, the ion beam from the mass analysis magnet apparatus 22 is deflected to a lower side to the side of the minus electrode 26-2 by the park electrode 26 as shown by a bold line in FIG. 5. In a case of an extracting voltage equal to or higher than several tens kV at the ion source 10, a voltage applied to the minus electrode 26-2 is preferably around 10% thereof, for example, around −10 kV. Evacuating of the ion beam by such a deflection is much faster (microsecond order) than that of a deflection which is carried out mechanically, and is referred to as high speed evacuating. The deflected ion beam maintains a state of impinging on a face on the upstream side of the mass analysis slit 28 or impinging on the graphite 122 of the inner wall of the park housing 27. The state normally continues for a short period of time of about several seconds and is referred to as temporary evacuating of the ion beam. When the potential difference is nullified by making the power source applied to the park electrode 26 off in such a state, there is brought about a state in which the electric field is not present at the park electrode 26, and the deflected ion beam recovers to be along the beam trajectory line indicated by a two-dotted chain line in FIG. 5. Thereby, an ion beam comprising an ion of a predetermined mass in the ion beam passing the park electrode 26 passes the hole 120 of the mass analysis slit 28. Further, the ion beam passing the hole 120 of the mass analysis slit 28 advances to the beam scanner 36 arranged on the downstream side. By providing the park electrode 26, even when the sectional shape of the ion beam is either of a normal circular shape and an elliptical or flat shape prolonged in a lateral direction (having long axis in lateral direction) or prolonged in the longitudinal direction (having long axis in longitudinal direction), the ion beam can excellently be evacuated at inside of the park housing 27 without being effected with an influence of the sectional shape of the ion beam. Further, a region on which the ion beam brought into a temporal evacuating state impinges is covered by the graphite member which is difficult to be sputtered by the ion beam, and therefore, an adverse influence of contamination or the like is not effected by sputtered particles on the downstream side of the mass analysis slit 28. Further, a magnetic field deflection may be adopted in place of an electric field deflection by the park electrode 26 as described above. Next, an explanation will be given of an operation when a desired ion beam cannot be provided by generating a discharge phenomenon at the ion implantation apparatus 1 constituted as described above, and a simple explanation will be given of ion implantation at normal time before the explanation. FIG. 6 is an explanatory view showing a behavior when an ion is implanted to the wafer 58. As shown by FIG. 6, a lifting apparatus 130 includes a platen (not illustrated) for holding the wafer 58, and moves up and down the wafer 58 by moving up and down the platen in the up and down direction. Further, the lifting apparatus 130 includes CPU (Central Processing Unit) 132 for carrying out control and RAM (Random Access Memory) 134 for storing a position information in the up and down direction of the wafer 58, and stores the position information of the wafer 58 as necessary. A pair of dose cups 140 are arranged at fixed positions at inside of a region of irradiating the ion beam, in this case, left and right positions of the lifting apparatus 130 for measuring a dose amount and outputting a measured value. A dose amount determination portion (resolving means) 142 determines whether the dose amount is proper based on the measured values from the pair of dose cups 140 and outputs a result of determination as a determining signal. Specifically, when the dose amount is equal to or larger than a predetermined value, the dose amount determination portion 142 outputs a determining signal indicating that the dose amount is proper (hereinafter, referred to as proper determining signal). On the other hand, when the dose amount is less than the predetermined value, the dose amount determination portion 142 outputs a determining signal indicating that the does amount is improper (hereinafter, referred to as improper determining signal). As indicated by an arrow mark of a broken line (arrow mark in lateral direction), the ion beam is reciprocally scanned to traverse the pair of dose cups 140 by the beam scanner 36. When the wafer 58 is moved in the up and down direction as indicated by an arrow mark of a bold line (arrow mark in up and down direction) relative to the ion beam reciprocally scanned in the horizontal direction, an entire face of the wafer 58 is scanned by the ion beam. As a result, the ion of the ion beam is implanted to an entire face of the wafer 58. Specifically, the ion is implanted to the entire face of the wafer 58 during a time period of moving the wafer 58 from a lowermost position to an uppermost position, or from the uppermost position to the lowermost position. Meanwhile, in a case in which a desired ion beam cannot be provided by a discharge phenomenon when the ion is implanted to the wafer 58 in this way, the dose amount measured by the dose cup 140 is reduced. Further, when the dose amount becomes less than the predetermined value, the dose amount determination portion 142 outputs an improper determining signal. When the improper determining signal is received, a park power source control portion (control means) 144 controls the park voltage to be applied to the park electrode 26. When the park voltage is applied, the park electrode 26 evacuates the ion beam instantaneously by deflecting the ion beam to the lower side from the beam trajectory line and the state is maintained for a predetermined time (for example, 2 seconds). As a result, the ion beam impinges on the graphite 122 of the mass analysis slit 28, or the graphite 122 at inside of the park housing 27. Therefore, the mass analysis slit 28 or the park housing 27 can be utilized as an evacuating location. In addition thereto, the ion beam is deflected to the graphite 122 at inside of the mass analysis slit 28 of the park housing 27, and therefore, the ion beam does not arrive at the wafer 58 and is not implanted to the wafer 58. Further, when the improper determining signal is received, CPU 132 of the lifting apparatus 130 stores information of the position in the up and down direction of the wafer 58 to RAM 134 and evacuates the wafer 58 to a position at which an ion is not implanted thereto (outside of region of irradiating ion beam) for caution's sake. Specifically, in a case in which a position of implanting an ion to the wafer 58 is an upper side than a center of the wafer 58 when the discharge phenomenon occurred, the wafer 58 is evacuated from the region of irradiating the ion beam by moving up the wafer 58 to the uppermost position. On the other hand, in a case in which the position of implanting the ion is a lower side than the center of the wafer 58 when the discharge phenomenon occurred, the wafer 58 is evacuated from the region of irradiating the ion beam by moving down the wafer 58 to the lowermost position. Therefrom, the lifting apparatus 130 and CPU 132 are operated as wafer evacuating means. Successively, when it is determined that the wafer 58 is evacuated to the uppermost position or the lowermost position by an elapse of a predetermined time period after receiving the improper determining signal, the park power source control portion 144 stops supplying the park voltage to the park electrode 26. As a result, the ion beam brought into the evacuated state recovers instantaneously to the beam trajectory line. When the ion beam recovers to the beam trajectory line, the periodical reciprocal scanning is carried out by the beam scanner 36, and therefore, the dose amount of the ion beam is measured by the dose cups 140. When the dose amount is equal to or larger than the predetermined value as a result of the measurement, the dose amount determination portion 142 outputs the proper determining signal. When the proper determining signal is received, the park power source control portion 144 applies the park voltage to the park electrode 26. When the park voltage is applied, the park electrode 26 evacuates the ion beam by instantaneously deflecting the ion beam to the lower side of the beam trajectory line. Further, when the proper determining signal is received, CPU 132 reads information of the position of the wafer 58 from RAM 134 and recovers the wafer 58 to a position when the discharge phenomenon is detected, by driving the lifting apparatus 130. Next, when it is determined that the predetermined time period has elapsed after receiving the proper determining signal and the wafer 58 is recovered to the position before evacuating (position when discharge phenomenon is detected), the park power source control portion 144 stops applying the park voltage to the park electrode 26. When the park voltage is stopped from being applied to the park electrode 26, the deflection of the ion beam is stopped and the ion beam instantaneously recovers to the beam trajectory line. As a result, the ion beam transmitting through the hole 120 of the mass analysis slit 28 advances to the beam scanner 36 and the ion beam is reciprocally scanned in the horizontal direction periodically by the beam scanner 36. At this occasion, the wafer 58 recovers to the position when the discharge phenomenon is detected, and therefore, ion implantation can be restarted from a midway position when ion implantation is interrupted. Therefore, even when the discharge phenomenon is brought about accidentally, so far as the phenomenon continues within the predetermined time period, it is not difficult to ensure the uniformity of the ion beam or makes the dose amount uniform and the ion can be implanted uniformly to the wafer 58. By providing the injector flag Faraday cup 32 by the above-described constitution, when the predetermined time period (for example, 2 seconds) has elapsed after receiving the improper determining signal indicating that the dose amount is improper and the dose amount is measured again, in a case in which the proper determining signal indicating that the dose amount is proper cannot be received, the ion beam is shut off by the injector flag Faraday cup 32 by advancing the injector flag Faraday cup 32 onto the beam trajectory line at the scanner housing 37. Naturally, temporary evacuating of the ion beam is released. By such an operation, the graphite 122 provided at the mass analysis slit 28 or the park housing 27 can be restrained from being sputtered without prolonging the time period of evacuating the ion beam by the deflecting apparatus. A control of the drive mechanism 32-1 can be realized by the park power source control portion 144. According to the combination of the park electrode 26 and the injector flag Faraday cup 32 as explained above, high speed evacuating and temporary evacuating of the ion beam can be realized without effecting an influence on the peripheral member without effecting an influence on the diameter or the sectional shape of the ion beam. Further, when a desired ion beam cannot be provided by combining dose amount measurement means, dose amount determination means, moving means (lifting means) as wafer evacuating means and CPU and RAM and park power source controlling means, by combining the temporal evacuating operation of the ion beam and the evacuating operation of the wafer, a nonuniform ion beam can be prevented from being irradiated to the wafer and the ion can be implanted uniformly always to the wafer. Further, the above-described embodiment can also be specified by being modified as follows. Although the ion implantation apparatus 1 according to the above embodiment is provided with the mass analysis magnet apparatus 22, the present invention is applicable also to an ion implantation apparatus which is not provided with the mass analysis magnet apparatus. This is a case of supplying a gas (for example, hydrogen or the like) which does not need to separate an ion by the mass analysis magnet apparatus. As shown by FIG. 7A and FIG. 7B, a two wire type beam profile monitor 31 for measuring a current density distribution of a section of the ion beam may be arranged to be able to be brought in and out to and from the beam trajectory line on an upstream side (FIG. 7A) or a downstream side (FIG. 7B) of an immediate vicinity of the injector flag Faraday cup 32. When constituted in this way, the quadrupole vertically focusing electromagnet 24 and the quadrupole vertically focusing electromagnet 30 can be adjusted based on the current density distribution measured by the two wire type beam profile monitor 31. Further, the quadrupole vertically focusing electromagnet 24 and the quadrupole vertically focusing electromagnet 30 can be adjusted also by comparing data measured by the triple surface beam dump 80 and data measured by the two wire type beam profile monitor 31. Although according to the embodiment, there is constructed a constitution of periodically scanning reciprocally the ion beam in the horizontal direction orthogonal to the direction of advancing the ion beam, instead thereof, there may be constructed a constitution of periodically scanning reciprocally the ion beam in a specific direction other than the horizontal direction, for example, a vertical direction. Although according to the embodiment, the present invention is applied to an ion implantation apparatus of a single wafer type, instead thereof, the present invention may be applied to an ion implantation apparatus of a batch type. |
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claims | 1. An apparatus to collimate an electromagnetic projection, the apparatus comprising:a collimator blade; andan edge having at least one physical property of varying ability to absorb electromagnetic energy, the edge being attached to the collimator blade;wherein cross section dimensions of the edge are described by a formula s=smax (exp(−μt)) wherein, s represents a signal, smax represents a maximum signal, μ represents an attenuation coefficient of the edge, and t represents a thickness of the edge. 2. The apparatus of claim 1, wherein the at least one physical property of varying ability to absorb electromagnetic energy further comprises:a tapered shape. 3. The apparatus of claim 2, wherein the at least one physical property of varying ability to absorb electromagnetic energy further comprises:a partially-squared tapered shape. 4. The apparatus of claim 1, wherein the at least one physical property of varying ability to absorb electromagnetic energy further comprises:a knife-edged shape. 5. The apparatus of claim 1, wherein the at least one physical property of varying ability to absorb electromagnetic energy further comprises:a varying composition of material in the edge. 6. The apparatus of claim 1, wherein the at least one physical property of varying ability to absorb electromagnetic energy further comprises:a plurality of materials having unequal atomic weights, in which the position of the materials is arranged in order from the material having the highest atomic weight being closest to the collimator blade, and the material having the lowest atomic weight being furthest from the collimator blade. 7. The apparatus of claim 6, wherein the plurality of materials further comprise:two materials. 8. The apparatus of claim 1, wherein the at least one physical property of varying ability to absorb electromagnetic energy further comprises:a varying density of material in the edge. 9. The apparatus of claim 1, wherein the at least one physical property of varying ability to absorb electromagnetic energy further comprises:a varying dimension of a shape of the edge. 10. An apparatus to collimate an electromagnetic projection, the apparatus comprising:a collimator blade; andan edge having a tapered shape, the edge consisting of a material, and the material having about uniform density throughout the edge, the edge being formed as part of the collimator blade;wherein cross section dimensions of the edge are described by a formula s=smax (exp(−μt)) wherein, s represents a signal, smax represents a maximum signal, μ represents an attenuation coefficient of the edge, and t represents a thickness of the edge. 11. The apparatus of claim 10, wherein the tapered shape further comprises:a knife-edge tapered shape. 12. The apparatus of claim 10, wherein the tapered shape further comprises:a partially-squared tapered shape. 13. An apparatus to collimate an electromagnetic projection, the apparatus comprising:an electromagnetic source; anda collimator having a tapered edge, the collimator being attached to the electromagnetic source;wherein cross section dimensions of the edge are described by a formula s=smax (exp(−μt)) wherein, S represents a signal, smax represents a maximum signal, μ represents an attenuation coefficient of the edge, and t represents a thickness of the edge. 14. The apparatus of claim 13, wherein the thickness of a leading edge is less than the thickness of an interior edge. 15. The apparatus of claim 14, wherein the thickness of the collimator at the leading edge of the collimator is about 10% of the thickness of the collimator at an interior position. 16. The apparatus of claim 13, wherein the tapered edge further comprises:a knife edge at the leading edge of the collimator. 17. An apparatus to limit a cross section of an X-Ray beam, the apparatus comprising:a source of the X-Ray beam; anda collimator having a tapered edge;wherein cross section dimensions of the tapered edge are described by a formula s=smax (exp(−μt)) wherein, s represents a signal, smax represents a maximum signal, μ represents an attenuation coefficient of the edge, and t represents a thickness of the edge. 18. The apparatus of claim 17, wherein the tapered edge further comprises:a thinner thickness at the inside edge of the collimator than at an interior position of the collimator. 19. The apparatus of claim 17, wherein the tapered edge further comprises:a knife edge at the inside edge of the collimator. 20. A medical X-Ray imaging system, the apparatus comprising:a source operable to project a medical X-Ray; anda collimator having an leading edge and an interior edge, the leading edge having a thickness that is less than a thickness of the interior edge;wherein cross section dimensions of the edge are described by a formula s=smax (exp(−μt)) wherein, s represents a signal, smax represents a maximum signal, μ represents an attenuation coefficient of the edge, and t represents a thickness of the edge. 21. The medical X-Ray imaging system of claim 20, the collimator further comprising:a tapered knife-edge shape. |
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047537736 | claims | 1. A heat exchanger for transferring heat from a primary fluid to a secondary fluid through an intermediate heat transfer fluid comprising a closed intermediate heat transfer fluid circuit comprising a disengaging chamber and a tubular coil portion connected at each end to said disengaging chamber, which coil portion is immersed in the primary fluid and wherein said intermediate heat transfer fluid circuit is partially filled with a non-circulating intermediate heat transfer fluid which is compatible with said primary fluid, and a secondary fluid circuit which passes through the disengaging chamber and through the tubular coil portion, and is substantially completely enclosed by said intermediate heat transfer fluid circuit, wherein the intermediate heat transfer fluid resides in the space enclosed within the tubular coil portion but outside said secondary fluid circuit and surrounds said secondary fluid circuit. each heat exchange module is comprised of at least one disengaging chamber, a multiplicity of double tube assemblies, at least one feedwater inlet and at least one steam outlet; each of said double tube assemblies is comprised of an inner tube individually enclosed for at least a portion of its length by an outer tube to form a double tube portion and thereby define an annular gap which is outside said inner tube but enclosed by said outer tube; each inner tube is attached at one end to a feedwater inlet, and each inner tube is attached at the other end to a steam outlet; said outer tube being in open communication at both ends with said disengaging chamber; said double tube portion being in the configuration of a coil for part of its length; said upper plenum has no communication with said disengaging chamber and has restricted communication with said lower plenum such that liquid metal entering the upper plenum and flowing to said lower plenum closely contacts at least a portion of the double tube assembiles; and said annular gap is at least partially filled with liquid metal. each heat exchange module is comprised at least one disengaging chamber, a multiplicity of double tube assemblies and a plurality of feedwater inlet nozzles and steam outlet nozzles, the number of feedwater inlet nozzles being equal to the number of steam outlet nozzles, and each of said nozzles providing communication to the outside of the cylindrical vessel; each of said double tube assemblies is comprised of an inner tube individually enclosed for at least a portion of its length by an outer tube to form a double tube portion and thereby define an annular gap which is outside said inner tube but enclosed by said outer tube; said inner tubes are attached at one end to a feedwater inlet and attached at the other end to a steam outlet nozzle; said outer tubes are in open communication at both ends with a disengaging chamber; said annular gap is at least partially filled with liquid metal; each double tube portion extends from its end closest to the feedwater inlet connection of its inner tube downwardly to the bottom of said upper plenum, then curves upwardly in a coil configuration for at least a portion of the length of said upper plenum, the remainder of said double tube portion extending upwardly to its end closest to the connection of its inner tube with a steam outlet nozzle; said upper plenum has at least one liquid metal inlet in open communication with the outside of the cylindrical vessel, said upper plenum having no communication with said disengaging chamber and having restricted communication with said lower plenum such that liquid metal entering the upper plenum and flowing downwardly to said lower plenum closely contacts at least a portion of the double tube helical coil; said lower plenum having at least one liquid metal outlet in open communication with the outside of the cylindrical vessel; said double tube assemblies are enclosed by a shroud extending the length of the upper plenum, the portion of said upper plenum outside said shroud being separated from said lower plenum by a diaphragm; the portion of said upper plenum outside said shroud is in communication with the portion enclosed by said shroud by means of a plurality of liquid metal distributor openings in said shroud. a vessel, a nuclear core which is cooled by a primary fluid, at least one heat exchanger for transferring heat from said primary fluid to a secondary fluid through an intermediate heat transfer fluid, said heat exchanger comprising a closed intermediate heat transfer fluid circuit comprising a disengaging chamber and a serpentine coil portion connected at each end to said disengaging chamber, which serpentine coil portion is immersed in the primary fluid and wherein said intermediate heat transfer fluid circuit is partially filled with a stagnant intermediate heat transfer fluid which is compatible with said primary fluid, and a secondary fluid circuit which passes through the vessel and is substantially completely enclosed by said intermediate heat transfer fluid circuit. a steam generator vessel having a closed lower end, divided into at least two longitudinally arranged sections including an upper plenum and a lower plenum, said upper plenum being above said lower plenum and containing a plurality of heat exchange modules, wherein each heat exchange module is comprised of at least one disengaging chamber, a multiplicity of double tube assemblies, at least one feedwater inlet and at least one steam outlet; each of said double tube assemblies is comprised of an inner tube individually enclosed for at least a portion of its length by an outer tube to form a double tube portion and thereby define an annular gap which is outside said inner tube but enclosed by said outer tube; each inner tube is attached at one end to a feedwater inlet, and each inner tube is attached at the other end to a steam outlet; said outer tube being in open communication at both ends with said disengaging chamber; said double tube portion being in the configuration of a coil for part of its length; said upper plenum has no communication with said disengaging chamber and has restricted communication with said lower plenum such that liquid metal entering the upper plenum and flowing to said lower plenum closely contacts at least a portion of the double tube assemblies; said annular gap is at least partially filled with stagnant liquid metal; and a guard vessel substantially completely enclosing said steam generator vessel, which guard vessel is equipped with vertical fins attached to the outer surface of the guard vessel and extending for at least a major portion of the length of the guard vessel, said fins providing a heat transfer surface for heat removal from the guard vessel and being capable of directing air flow vertically along the surface of said guard vessel and its fins; said method comprising: (1) circulating water to the steam generator and condensing the steam in a condenser, or (2) connecting one or more inner tubes to a cooling tower whereby the steam generated in the inner tubes is condensed in the cooling tower and recycled to the inner tubes, or (3) circulating air under the guard vessel such that cooling air is channeled along the sides of the guard vessel by the vertical fins, or (4) any combination of (1), (2) or (3), above. 2. A heat exchanger as defined in claim 1, wherein said intermediate heat transfer fluid circuit includes a multiplicity of coil portions. 3. A heat exchanger as defined in claim 2, wherein said intermediate heat transfer fluid circuit has a removeable modular configuration. 4. A heat exchanger as defined in claim 3, wherein said disengaging chamber includes means for detecting a rupture in said intermediate heat transfer fluid circuit or said secondary fluid circuit. 5. A heat exchanger as defined in claim 3, wherein said coil portions are comprised of concentric double tube assemblies, wherein the secondary fluid circuit comprises the inner tube and the intermediate heat transfer fluid circuit comprises the outer tube exclusive of the inner tube. 6. A heat exchanger as defined in claim 4 which further includes a circulation pump located centrally in a separate core module, which pump circulates cooled primary fluid back to a heat source. 7. A steam generator comprising a container having a closed lower end, divided into longitudinally arranged sections including an upper plenum and a lower plenum, said upper plenum being above said lower plenum and accomodating at least one removeable heat exchange module, wherein 8. A steam generator as defined in claim 7, wherein each heat exchange module is suspended from a support grid connected to said container near its upper end. 9. A steam generator as defined in claim 8, wherein said support grid forms a plurality of cavities providing seats from which a corresponding plurality of heat exchange modules may be supported. 10. A steam generator as defined in claim 9, wherein said support grid forms five square cavities of equal size in a cruciform configuration. 11. A steam generator as defined in claim 10, which further includes a core module supported in the central cavity formed by said support grid and wherein four heat exchange modules are supported in the outer four cavities. 12. A steam generator as defined in claim 11, wherein said core module contains a pump with intake means in communication with said lower plenum. 13. A modular steam generator comprising a cylindrical vessel having a closed lower end, divided into at least two longitudinally arranged sections including an upper plenum and a lower plenum, said upper plenum being above said lower plenum and containing support means for at least one removeable heat exchange module, wherein 14. A steam generator as defined in claim 13, wherein said coil configuration is a serpentine coil configuration. 15. A steam generator as defined in claim 14, wherein each heat exchange module includes 10-1000 double tube assemblies and each nozzle is connected to 10-200 inner tubes. 16. A steam generator as defined in claim 14, wherein the support means comprises a grid of straight members connected to the cylindrical vessel near its upper end, said grid forming a cruciform pattern of five equally sized square cavities for receiving heat exchange and equipment modules. 17. A steam generator as defined in claim 16, wherein the cylindrical vessel is substantially completely enclosed in a guard vessel. 18. A steam generator as defined in claim 16, wherein said inner tubes have an outside diameter of about 1.25 inches and said outer tubes have an inside diameter of about 1.615 inches and an outside diameter of about 1.75 inches. 19. A steam generator as defined in claim 16, wherein a spacer is provided between said inner and outer tubes. 20. A steam generator as defined in claim 16, wherein said liquid metal in said annular gap is sodium or a sodium-potassium mixture. 21. A steam generator as defined in claim 16, further including a jet eductor located in said lower plenum having intake means positioned to receive both liquid metal flowing from the upper plenum and liquid metal being discharged by a discharge pump, housed in a core module located in the central cavity of said support grid, said pump having intake means in communication with said lower plenum and directing its discharge through said liquid metal outlet. 22. A steam generator as defined in claim 16, wherein the double tube assemblies are fabricated from a low alloy steel selected from 21/4Cr - 1 Mo or 9 Cr - 1 Mo, and the cylindrical vessel is fabricated from high alloy steel selected from 304 SS or 316 SS, and the structural connections wherein low alloy steel and high alloy steal are joined are accomplished without a bimetallic weld. 23. A steam generator as defined in claim 16, wherein detection means are in communication with said disengaging chamber which are capable of detecting failure of an individual inner tube within a double tube assembly or failure of an individual outer tube. 24. A steam generator as defined in claim 23, wherein said detection means for failure of an inner tube include a hydrogen detector probe in the disengaging chamber and wherein said detection means for failure of an outer tube include a liquid level or temperature probe monitoring the height or temperature of liquid metal in the double tube portion of said helical coils. 25. A steam generator as defined in claim 23, further including blow-out seals in communication with the volume of the disengaging chamber which will rupture at the increase in pressure within the disengaging chamber caused by the reaction in one double tube portion between the liquid metal in the annular gap and water leaking from a failed inner tube in said double tube portion or caused by leakage from a water or steam tube within the disengaging chamber. 26. A steam generator as defined in claim 23, wherein said disengaging chamber is in communication with a purge line and a drain line, each equipped with a valve closure, said drain line providing communication between the disengaging chamber and disposal means for solid or liquid material entering the disengaging chamber. 27. A steam generator as defined in claim 23, which further includes one or more gas seals between the diaphragm and the shroud such that when the seals are breached, liquid metal entering the upper plenum may flow directly to the lower plenum, and wherein said gas seals and said liquid metal distributor openings provide the only means of communication between the upper plenum and the lower plenum. 28. A steam generator as defined in claim 27, wherein said cylindrical vessel is substantially completely enclosed in a guard vessel, which guard vessel is equipped with vertical fins attached to the outer surface of the guard vessel and extending for at least a major portion of the length of the guard vessel, said fins providing a heat transfer surface providing heat removal from the guard vessel and being capable of directing air flow vertically along the surface of said guard vessel and its fins. 29. A steam generator as defined in claim 28, wherein a layer of insulating material surrounds the guard vessel, supported at the ends of said vertical fins. 30. A steam generator as defined in claim 28, wherein said disengaging chamber is subdivided into discrete sections, each section corresponding to an inlet or outlet nozzle and enclosing a separate bundle of tubes, such that the double tube portion annular gaps of each tube bundle are in communication with only one disengaging chamber section at either end of the double tube portion, and such that isolation of an individual double tube bundle by closing off its inlet and outlet nozzles, or isolation of any individual inner tube by sealing its inlet and outlet opening, does not influence the operation of the rest of the steam generator. 31. A steam generator as defined in claim 30, wherein vent and drain holes interconnect each inlet disengaging chamber and separate vent and drain holes interconnect each outlet disengaging chamber, and a rupture disc and a fill/drain line connect to each disengaging chamber. 32. A nuclear power plant comprising a nuclear reactor having a circulating liquid metal cooling system, which cooling system is connected to the steam generator as defined in claim 29. 33. A pool reactor comprising 34. A pool reactor as defined in claim 33, having a circulation pump immersed in the primary fluid which returns primary fluid discharged from the heat exchanger to the nuclear core. 35. A method for removing decay heat in a nuclear power plant comprising a nuclear reactor having a circulating liquid metal cooling system, which cooling system includes at least one steam generator comprising |
summary | ||
claims | 1. An electrostatic lens array comprising:multiple substrates spaced from each other by intervals, each of the multiple substrates having an aperture for passing therethrough a charged particle beam,wherein with respect to a travelling direction of the charged particle beam, a peripheral contour line formed by any one of surfaces of the multiple substrates other than an upper surface of a most upstream substrate and a lower surface of a most downstream substrate has a protruding portion protruding from a peripheral contour line of both of the upper surface of the most upstream substrate and the lower surface of the most downstream substrate,wherein a position of the protruding portion is defined by a position regulating member, whereby parallelism is adjustable so that a surface including the protruding portion is parallel to a surface to be irradiated with the charged particle beam after passing through the aperture, andwherein the position regulating member is in contact with the protruding portion. 2. The electrostatic lens array according to claim 1, further comprising a support portion configured to support the protruding portion, wherein the parallelism is adjustable by the position regulating member via the support portion. 3. The electrostatic lens array according to claim 1, wherein the position regulating member comprises any one of a laser measuring machine, a capacitive sensor, an abutting jig, and a combination thereof. 4. The electrostatic lens array according to claim 1, wherein the multiple substrates comprise three electrodes in which a negative voltage is applied to an intermediate electrode, and other electrodes are connected to a ground. 5. A charged particle optical system comprising:a collimator lens configured to collimate a charged particle beam emitted from a charged particle source;an aperture array configured to divide the collimated charged particle beam; andan electrostatic lens array configured to focus each of the divided charged particle beams so as to irradiate a sample surface,wherein the electrostatic lens array includes multiple substrates spaced from each other by intervals, each of the multiple substrates having an aperture for passing therethrough the charged particle beam,wherein with respect to a travelling direction of the charged particle beam, a peripheral contour line formed by any one of surfaces of the multiple substrates other than an upper surface of a most upstream substrate and a lower surface of a most downstream substrate has a protruding portion protruding from a peripheral contour line of both of the upper surface of the most upstream substrate and the lower surface of the most downstream substrate;wherein a position of the protruding portion is defined by a position regulating member, whereby parallelism is adjustable so that a surface including the protruding portion is parallel to a surface to be irradiated with the charged particle beam after passing through the aperture, andwherein the position regulating member is in contact with the protruding portion. 6. The charged particle optical system according to claim 5, further comprising a blanker array configured to deflect each of the divided charged particle beams in accordance with a drawing pattern. 7. An electrostatic lens array comprising:multiple substrates including at least first and second substrates; anda spacer contacting a lower surface of the first substrate and an upper surface of the second substrate to space to the first and second substrates from each other, each of the multiple substrates having an aperture for passing therethrough a charged particle beam,wherein with respect to a travelling direction of the charged particle beam, the spacer has a protruding portion protruding from a peripheral contour line of one of the upper surface of the most upstream substrate and the lower surface of the most downstream substrate,wherein a position of the protruding portion is defined by a position regulating member, whereby parallelism is adjustable so that a surface including the protruding portion is parallel to a surface to be irradiated with the charged particle beam after passing through the aperture, andwherein the position regulating member is in contact with the protruding portion. |
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description | This application claims the benefit of priority to U.S. Provisional Patent Application Ser. No. 62/630,00 filed Feb. 13, 2018, which is incorporated by reference in its' entirety. This application relates to cleaning of nuclear waste storage tanks containing radioactive liquids, solids, and sludge waste, where these tanks, approaching approximately 100 feet in diameter, have only an approximately 12-inch diameter opening to pass through, and in particular to systems, devices, and methods for passing vertical reach robotic tank cleaning systems and devices through the opening and mounted to existing riser structure. The robotic vertical reach tank cleaning systems and devices can extend to the bottom of the tank and horizontally more than 30 feet. The systems and devices are light weight and can be supported by the existing structure of the tank. Two or more of these systems and devices can be used within a single tank to break down the solid sludge by placing the high-pressure nozzle in close proximity to the solid waste and providing maximum coverage. A mechanical arm with a nozzle assembly utilizes high- and low-pressure fluid streams to fluidize solids while directly motivating them in the direction of a centrally located transfer pump. The robotic vertical reach tank cleaning systems, devices and methods can work using hydraulic actuation in highly radioactive, chemically aggressive, explosive, environments with high temperature conditions up to approximately 212 degrees Fahrenheit, or low temperature conditions down to approximately 32 degrees Fahrenheit. Multiple axis of freedom allows the arm and nozzle assembly to navigate and clean around internal tank obstacles and reach close proximity for maximum nozzle impingement force. The robotic vertical reach tank cleaning systems, devices and methods can be remotely operated up to 1000 feet away. The various capabilities of this invention alleviate the difficulties of access to nuclear storage tanks with greater depths of solid waste. At nuclear waste storage facilities, radioactive material is generally stored in underground tanks. Historically, cleaning of the underground tanks has typically been done in conical bottom tanks that allow liquid to drain towards a centrally located transfer pump. With flat bottom tanks, the liquid is not pulled by gravity towards the tank center. As such, it is more difficult to direct flow to the pump; and areas behind obstructions can become difficult to clean efficiently or effectively. A vertical reach cleaning system can position the nozzle(s) such that the water jet directs the liquefied waste towards the transfer pump. However, waste depth in some tanks can be too high to allow a long arm cleaning system of adequate length to be installed. Short arm cleaning systems have been considered, but this limits the high-pressure nozzle proximity to the waste during the majority of the cleaning of the tank, thus, reduces impingement force and the range of the cleaning system, and limits the capability to clean obstructed areas. The present invention seeks to provide a solution that solves the above challenges and provides functional cleaning capabilities in tanks with large waste build up and obstructions. A primary objective of the present invention is to provide systems, devices and methods of providing a vertically adjustable robotic tank cleaning system for use in nuclear waste storage tanks. A vertical travel robotic tank cleaning system can be mounted to existing pipe tank risers as small as approximately 12 inches in diameter. Nozzles mounted on the end of telescoping arms can utilize fluid jets to break up, liquefy and motivate solids. The vertical travel robotic tank cleaning system can have up to approximately 6 degrees of freedom or more. The arm mast section rotates the mast, boom, and nozzle assemblies +/− approximately 180 degrees about the vertical longitudinal axis. A telescoping boom allows the extension and retraction of the nozzles over approximately 30 feet to maneuver around and within the perimeter of the tank. The boom can rotate over approximately 90 degrees from vertical to horizontal. The nozzle assembly at the distal end of the boom can be twisted and rotated to direct the liquid stream as needed. The vertical adjustability of the mast allows for the installation of longer arms into waste tanks with high waste depths. When the vertical travel robotic tank cleaning system is initially installed in a tank with a lot of solids, the nozzle(s) first erodes material from below the arm. As the vertical travel robotic tank cleaning system mast is incrementally lowered, more waste can be cleared from below the device. Once the vertical travel robotic tank cleaning system removes enough waste the boom and nozzle assembly can be rotated to elevate the components above the waste level. At this point, the long arm boom can be extended to be within close proximity of a large portion of the tank. With the extended reach of the boom and the requisite nozzle orientation, great access to tank and areas shadowed by obstructions can be reached in such a way that the stream of liquefied waste can be directed towards the transfer pump. The nozzle stream can be directed into, behind, above, and below obstructions such as air lift circulators. The vertical adjustability allows the low flow high pressure orbital wash nozzles to have a closer proximity to the waste and larger effective working area by getting the nozzle closer to the work across a larger area. The nozzle assembly can utilize a single or multiple low pressure, up to and beyond approximately 500 psi, high flow, up to approximately 500 gpm, nozzle(s), and either one, a pair or more of integrated high-pressure low flow orbital wash nozzles that provide a solid, zero-degree water fluid stream working at up to and above approximately 5,000 pounds per square inch of water pressure and flowrates up to approximately 50 gpm. The nozzle rotates the water stream in a conical pattern up to approximately 25 degrees included angle from the tip of the nozzle outlet. The vertical travel robotic tank cleaning system is operated from a remote console station up to approximately 1000 feet away. It is hydraulically driven through a hydraulic manifold utilizing adjustable valves located directly outside the tank. Further objects and advantages of this invention will be apparent from the following detailed description of the presently preferred embodiments which are illustrated schematically in the accompanying drawings. Before explaining the disclosed embodiments of the present invention in detail it is to be understood that the invention is not limited in its applications to the details of the particular arrangements shown since the invention is capable of other embodiments. Also, the terminology used herein is for the purpose of description and not of limitation. In the Summary above and in the Detailed Description of Preferred Embodiments and in the accompanying drawings, reference is made to particular features (including method steps) of the invention. It is to be understood that the disclosure of the invention in this specification does not include all possible combinations of such particular features. For example, where a particular feature is disclosed in the context of a particular aspect or embodiment of the invention, that feature can also be used, to the extent possible, in combination with and/or in the context of other particular aspects and embodiments of the invention, and in the invention generally. In this section, some embodiments of the invention will be described more fully with reference to the accompanying drawings, in which preferred embodiments of the invention are shown. This invention may, however, be embodied in many different forms and should not be construed as limited to the embodiments set forth herein. Rather, these embodiments are provided so that this disclosure will be thorough and complete, and will convey the scope of the invention to those skilled in the art. Like numbers refer to like elements throughout, and prime notation is used to indicate similar elements in alternative embodiments. A list of components will now be described. A First embodiment vertical travel robotic tank cleaning system 1 primary enclosure 2 support tube 3 rectangular mast 4/22 telescopic boom section(arms) 5 pivot point 5HC boom elevation hydraulic cylinder 6 rigid nozzle assembly N6 nozzles with pan & tilt OWN orbital wash nozzles 6B base of nozzle assembly 6AX perpendicular axis 7 adapter spool 8 outer circular tube section 9 inner circular tube section 10 slide pads 11 outside rectangular mast tube 12 inside rectangular mast tube 13 slide pads 14 process hose 15 hydraulic cylinders 16 hydraulic hose 17 bushing/bearing 18 hose management systems B Second embodiment vertical travel robotic tank cleaning system with rack & pinion 19 primary enclosure 20 large diameter support tube 21 rectangular mast 22 telescopic boom sections 23 pivot point 24 nozzle assembly 25 hose management system 26 hose reel/adapter spool 27 high reduction gear box 28 pinion gear 29 rack gear 30 mast support tube 31 mating key 32 key seat 33 adapter spool 34 hydraulic motor C Third embodiment vertical travel robotic tank cleaning system with chain & sprocket(s) 40 chain 41 sprocket(s) 42 idler sprocket(s) 50 turntable 60 manifold system 70 nozzle assembly pivot motor 72 upper sprocket 74 lower sprocket 76 chain 80 nozzle assembly rotate motor 82 gears 84 nozzle assembly rotation base 86 nozzle assembly rotation stem 100 generic waste tank A first embodiment A, of the vertical travel robotic tank cleaning system is a hydraulically actuated, telescopic, functional in hazardous/explosive environments, able to fit through risers as small as approximately 12 inches in diameter and further extends the reach of the telescopic boom. FIG. 1A shows a retracted (left) view of one embodiment A of the vertical travel robotic tank cleaning system. FIG. 1B shows an extended (right) view of the embodiment A of the vertical travel robotic tank cleaning system of FIG. 1A. Referring to FIGS. 1A-1B and 1C, the vertical travel robotic tank cleaning system A can comprise a primary enclosure 1 that contains a turntable 50 and hose management assembly that accommodates axial and radial motion of the arm. Extending down from the turntable assembly 50, a rectangular mast 3 can be rotated about the longitudinal axis within a large diameter support tube 2/20. Elevation of the telescopic boom sections 4 can be about the pivot point 5 with up to approximately 90 degrees of rotation to be perpendicular to mast 3. Referring to FIG. 4B, an inner end of telescopic boom arms 4/22 pivots to mast 3 by a pivot point 5/23 from a longitudinal position to approximately 90 degrees of rotation by an extendable and retractable boom elevation hydraulic cylinder 5HC. Turntable 50 allows for rotation of mast 3 with perpendicular boom arms 4/22. The nozzle assembly 6, employing both low pressure, up to approximately 500 psi and high flow up to approximately 500 gpm nozzle(s) N6, as well as high pressure, low flow orbital wash nozzle(s) OW6, that provide a solid, zero-degree water fluid stream rotating in a conical pattern up to a 25 degree included angle from the tip of the nozzle outlet at pressures up to approximately 5,000 pounds per square inch of water pressure and flowrates up to approximately 50 gpm. The nozzle assembly moves through two degrees of freedom allowing it to twist about the longitudinal axis and rotate about a perpendicular axis. Nozzle assembly 6 includes a longitudinal axis drive motor 80 which rotates gears 82 causing a rotation stem 86 extending below a rotation base 84 to rotate to twist about a longitudinal axis. A nozzle assembly pivot motor 70 with a perpendicular axis rotates an upper sprocket 72, which causes a chain 76 to rotate a lower sprocket 74, which causes a rigid base 6B of the rigid nozzle assembly 6 to pivot and rotate about a perpendicular axis. Feeding the nozzle assembly 6 and running along the inside of the rectangular mast 3 and telescopic boom sections 4 is the process hose that leads to a hose reel in the primary enclosure 1. A hose management system can be used to control the additional length of hydraulic hose running down the mast 3. An adapter spool 7 can support the vertical travel robotic tank cleaning system and interfaces with the tank. FIG. 2 illustrates a longitudinal cross section view of the embodiment shown in FIGS. 1A and 1B. FIG. 3 illustrates a cross section view of the embodiment as FIGS. 1A, 1B and perpendicular to FIG. 2. Referring to FIG. 2 and FIG. 3, this embodiment of the vertical travel robotic tank cleaning system can utilize a telescopic mast assembly and can be comprised of an outer circular tube section 8 and an inner circular tube section 9 that can slide past each other lengthwise with slide pads 10 therebetween. A bushing or a bearing 17 cam facilitate the twisting of the mast assembly within the large diameter tubes. Along the center axis of the circular tubes can be an outside rectangular mast tube 11 and an inside rectangular mast tube 12 that slide past each other lengthwise with slide pads 13 therebetween. Running along the inside of the inner most tube can be the process hose 14 that leads to the hose reel in the primary enclosure. In the preferred embodiment, hydraulic cylinders 15 can be used to provide the force for extending and retracting. In a further embodiment, a rack and pinion gearset can be used to extend and retract the mast. Alternately, a chain and sprocket could be used. In an even further embodiment, a lead screw assembly could be used to extend and retract the mast. These embodiments could be hydraulic, pneumatic, or electrical motor driven. A hose management system 18 can guide the hydraulic hose 16 exiting the inner rectangular mast tube as it extends or retracts. In reference to FIG. 4A, the vertical travel robotic tank cleaning system A of the previous FIGURES is shown retracted and installed in a generic waste tank 100 with a high waste depth. In reference to FIG. 4B, the vertical travel robotic tank cleaning system A of the previous FIGURES is shown extended and installed in the same waste tank 100 as FIG. 4A with the waste partially eroded to allow extension and elevation. A second embodiment of the vertical travel robotic tank cleaning system B can be hydraulically driven, utilizes a rack and pinion, functional in hazardous/explosive environments, able to fit through risers as small as approximately 12 inches in diameter and further extends the reach of the telescopic boom. FIG. 5A illustrates a retracted (left) view of a second embodiment B of the vertical travel robotic tank cleaning system. FIG. 5B illustrates an extended (right) view of the second embodiment of the vertical travel robotic tank cleaning system B of FIG. 5A. FIG. 5C is an enlarged view of a portion of the extended vertical travel robotic tank cleaning system B of FIG. 5B illustrating the adapter spool 33 supporting a pinion gear 28 and high reduction gearbox 27 that drives a rack gear 29 mounted to the large diameter support tube 20 vertically up or down. FIG. 5D is a view of the primary enclosure of FIGS. 1A, 1B 5A-5B with the covers removed. Referring to FIGS. 1A, 1B, 5A, 5B, 5C, and 5D, the vertical travel robotic tank cleaning system B can comprise a primary enclosure 19 that contains a turntable and hose management assembly that accommodates axial and radial motion of the arm. FIG. 5D shows a turntable 50 which can rotate arm 5 in FIG. 1A, 1B, and a manifold system 60 for supplying and returning working hydraulic fluid. Extending down from the turntable assembly, a rectangular mast 21 can be rotated about the longitudinal axis within a large diameter support tube 20. The elevation of the telescopic boom sections 22 can be about the pivot point 23 with up to approximately 90 degrees of rotation. Referring to FIGS. 5A, 5B, 5E and 5F, the nozzle assembly 24 can employ both low pressures, up to approximately 500 psi, high flow, up to approximately 500 gpm nozzles N6, and high-pressure low flow orbital wash nozzle(s) OW6 with solid, zero-degree fluid steam(s) working at up to, above and including 5,000 pounds per square inch of water pressure, and flowrates up to approximately 50 gpm The nozzle assembly 24 moves through two degrees of freedom allowing it to twist about the longitudinal axis and rotate about a perpendicular axis. Feeding the nozzle assembly 24, and running along the inside of the rectangular mast 21 and telescopic boom sections 22 can be the process hose that leads to a hose reel 26 in the primary enclosure 19. Along the telescopic boom section 22, a hose management system 25 can guide and support the hydraulic hose during extension and retraction. An adapter spool 26 can be used to support and interface the vertical travel robotic tank cleaning system B to the tank 100. FIG. 6 illustrates a longitudinal cross section view of the adapter spool assembly 33 of the second embodiment B as FIGS. 5A-5D. FIG. 7 illustrates a cross sectional view of the adapter spool assembly 33 of the second embodiment B as FIGS. 5A-5D. Referring to FIG. 6 and FIG. 7, the vertical travel robotic tank cleaning system B can be raised and lowered through a rack gear 29 and pinion 28 gear set. The rack gear 29 can be mounted longitudinally about the outer perimeter of the mast support tube 30. Clocked 180 degrees from the rack gear 29 is a key seat 32. The key seat 32 can be mounted longitudinally about the outer perimeter of the outermost mast support tube 30. The outer support tube 30 can fit concentrically through an adapter spool 33 with a mating key 31 to provide an anti-rotation feature. The mating key 31 is stationary and rigidly affixed to the adapter spool 33. The key has a protruding feature that is matched to the internal feature of the key seat 32, but the width is slightly undersized, ranging from approximately 0.0005″ to approximately 0.0625″ smaller allowing it to slide longitudinally through the stationary key seat; however, the shape of the key 31 and key seat 33 prevent rotation about the longitudinal axis. The key 31 can be mounted longitudinally about the outer perimeter of the adapter spool 33. Also, on the adapter spool 33 can be a pinion 28 supported between two bearings and driven by a hydraulic motor 34 through a high reduction gearbox 27. Driving with the rack gear 29, the pinion rotates either clockwise or counterclockwise to raise or lower the arm. FIG. 8A illustrates a cross sectional view of a waste tank 100 and the second embodiment B shown in FIGS. 5A-5D installed in a typical riser and lowered to a starting position just above the waste. FIG. 8B illustrates a cross sectional view of the waste tank 100 as the second embodiment B of FIGS. 5A-5D is lowered and the boom pivoted into the horizontal position. Referring to FIG. 8A, the multi axis distal nozzle cleaning arm B is shown retracted and installed in a generic waste tank 100 with a high waste depth. Referring to FIG. 8B, the multi axis distal nozzle cleaning arm B is shown extended and installed in the same waste tank 100 as FIG. 8A with the waste partially eroded to allow extension and elevation. FIG. 9A illustrates a retracted (left) view of a third embodiment C of the vertical travel robotic tank cleaning system. FIG. 9B illustrates an extended (right) view of the third embodiment C of the vertical travel robotic tank cleaning system of FIG. 9B. FIG. 9C is an enlarged view of a portion of the extended vertical travel robotic tank cleaning system of FIG. 9B illustrating the adapter spool 33 supporting a sprocket 41 and high reduction gearbox 27 that drives a chain 40 affixed to the large diameter support tube 20 vertically up or down. A third embodiment of the vertical travel robotic tank cleaning system C can be hydraulically driven, utilizes a chain and sprocket, functional in hazardous/explosive environments, able to fit through risers as small as 12 inches in diameter and further extends the reach of the telescopic boom. Referring to FIGS. 9A, 9B and 9C, the vertical travel robotic tank cleaning system C can include a primary enclosure 19 that contains a turntable and hose management assembly that accommodates axial and radial motion of the arm. Extending down from the turntable assembly, a rectangular mast 21 can be rotated about the longitudinal axis within a large diameter support tube 20. Elevation of the telescopic boom sections 22 can be about the pivot point 23 with up to approximately 90 degrees of rotation. The nozzle assembly 24, employing both low pressure, up to approximately 500 psi, high flow, up to approximately 500 gpm, and high pressure low flow nozzle(s) orbital wash nozzle(s) that provide a solid, zero-degree water fluid stream rotating in a conical pattern up to a 25 degree included angle from the tip of the nozzle outlet at pressures up to, above and including 5,000 pounds per square inch of water pressure and flowrates up to approximately 50 gpm. The nozzle assembly moves through approximately two degrees of freedom allowing it to twist about the longitudinal axis and rotate about a perpendicular axis. Feeding the nozzle assembly 24, and running along the inside of the rectangular mast 21 and telescopic boom sections 22 can be the process hose that leads to a hose reel 26 in the primary enclosure 19. Along the telescopic boom section 22, a hose management system 25 guides and supports the hydraulic hose during extension and retraction. The adapter spool 26 supports and interfaces the vertical travel robotic tank cleaning system to the tank 100 (shown in previous FIGURES. FIG. 10 illustrates a longitudinal cross section view of the adapter spool assembly 33 of the embodiment C shown in FIGS. 9A-9C. FIG. 11 illustrates a cross section view of the adapter spool assembly 33 of the of the embodiment C shown in FIGS. 9A-9C. Referring to FIG. 10 and FIG. 11, the vertical travel robotic tank cleaning system C can be raised and lowered through a chain 40 and driven sprocket 41 set. The chain 40 can be mounted longitudinally about the outer perimeter of the mast support tube 30. Clocked approximately 180 degrees from the chain 40 can be a key seat 32. The key seat 32 can be mounted longitudinally about the outer perimeter of the outermost mast support tube 30. The outer support tube 30 fits concentrically through an adapter spool 33 with a mating key 31 to provide an anti-rotation feature. The key 31 can be mounted longitudinally about the outer perimeter of the adapter spool 33. The mating key 31 is stationary and rigidly affixed to the adapter spool 33. The key has a protruding feature that is matched to the internal feature of the key seat 32, but the width is slightly undersized, ranging from approximately 0.0005″ to approximately 0.0625″ smaller allowing it to slide longitudinally through the stationary key seat; however, the shape of the key 31 and key seat 33 prevent rotation about the longitudinal axis. Also, on the adapter spool 33 can be a sprocket 41 supported between two bearings and driven by a hydraulic motor 34 through a high reduction gearbox 27. Driving the chain 40, the sprocket 41 rotates either clockwise or counterclockwise to raise or lower the arm. On each side of the driven sprocket can be idler sprockets 42. The systems, devices and methods can be remotely and robotically controlled with an interface which can allow a remotely located human operator to monitor and manipulate the process in real-time using controls such as but not limited to joysticks, and the like, as described in U.S. patent application Ser. No. 15/854,424 filed Dec. 26, 2017 entitled: Articulating Arm Programmable Tank Cleaning Nozzle, to the same assignee as the subject patent application, and which is incorporated by reference in its' entirety. The term “approximately” can be +/−10% of the amount referenced. Additionally, preferred amounts and ranges can include the amounts and ranges referenced without the prefix of being approximately. While the invention has been described, disclosed, illustrated and shown in various terms of certain embodiments or modifications which it has presumed in practice, the scope of the invention is not intended to be, nor should it be deemed to be, limited thereby and such other modifications or embodiments as may be suggested by the teachings herein are particularly reserved especially as they fall within the breadth and scope of the claims here appended. |
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abstract | In a proximity X-ray exposure apparatus using a point source as a light source, a laser beam (21) is focused on a circular disk-like target (14) arranged in an X-ray source unit to irradiate it, thus generating a plasma and then X-rays (17). The circular disk-like target (14) has four types of quadrant first-, second-, third-, and fourth-wavelength light-emitting portions (14a, 14b, 14c, 14d) which are made of different materials and divided equiangularly in the radial direction. The target (14) is rotatably controlled by a rotational drive mechanism in synchronism with pulse emission. As the rotational angle of the target (14) is controlled in synchronism with emission of the laser beam (21), the material (type) of the target (14) can be selected from the first-, second-, third-, and fourth-wavelength light-emitting portions (14a, 14b, 14c, 14d). |
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053405064 | summary | BACKGROUND OF THE INVENTION This invention relates to a method for immobilizing radioactive wastes for permanent disposal. More particularly, the invention relates to a method of immobilizing mixed waste chloride salts containing radionuclides and other hazardous materials for permanent disposal. The recovery of fissionable materials such as uranium and plutonium from spent nuclear reactor fuels can be carried out by electrorefining methods using electrochemical cells of the type described in U.S. Pat. Nos. 4,596,647 and 2,951,793, as well as U.S. Pat. No. 4,880,506. It is the electrorefining method which is being developed for the reprocessing of Integral Fast Reactor (IFR) fuel. In a typical electrorefining cell an electrolyte consisting of a molten eutectic salt mixture such as KCl and LiCl is used to transport the metal or metals to be purified between electrode solutions. When used to reprocess spent nuclear reactor fuels, the salt mixture becomes contaminated with radionuclides, such as .sup.137 cesium, .sup.90 strontium and .sup.129 iodine, hazardous materials such as barium and other species such as sodium, and eventually is no longer suitable for use in the electrorefining cell. Ideally, the salt would be decontaminated by removing a fraction of the heat-producing radionuclides, primarily cesium and strontium, and any other elements e.g. barium and sodium, which could potentially interfere in the operation of the electrorefiner, and the purified salt would be recycled back to the electrorefiner. However, the separation of cesium and strontium from the salt is difficult, and if they are separated in concentrated form, it would be necessary to dilute them in another matrix material and/or cool them before they could be stored since they are large heat producers. It is, therefore, more practical to dispose of the cesium and strontium and any other radionuclides, including iodides, and toxic metal chlorides along with a portion of the salt matrix. The waste salt containing the cesium, strontium and iodine is a high level waste (HLW), and as such must be disposed of in the geologic repository for HLW. This requires that the waste form be leach resistant to prevent an uncontrolled release of the radionuclides and other hazardous chemicals such as barium, into the groundwater. Since waste salts are chlorides and are very water soluble, a method for encapsulating and immobilizing the waste salt must be identified. One problem with developing a waste storage medium is that the waste salt consists primarily of chloride salts of alkali metals and as such is not readily amenable to treatment using procedures and techniques developed for immobilizing cesium and strontium in other nuclear waste streams. For instance, the chloride salts cannot be added directly to glass-forming compounds and processed to yield a leach-resistant glass since glasses containing halide ions are relatively water soluble. Therefore, for immobilization in a glass matrix the waste chloride salts must be converted into oxides or other chemical forms compatible with the glass-making process. However, conversion processes are expensive and time-consuming and raise environmental concerns about off-gases produced by the processes. A mortar matrix has also been considered as a possible waste form for the waste chloride salt. A special mortar was developed to incorporate lithium, potassium, cesium and strontium chloride salts into its structure, thereby immobilizing them. However, when irradiated, the water in the mortar was radiolyzed and hydrogen gas was generated. U.S. patent application Ser. No. 744,753, filed Aug. 14, 1991, and incorporated herein by reference, describes the use of certain zeolites to decontaminate and immobilize waste chloride salts. Contact between the zeolite (for example, zeolite A or mixtures of chabazite and erionite zeolites or mixtures thereof) in the sodium, potassium, or lithium form and the molten salt result in an ion exchange between the radionuclides cesium and strontium and the hazardous material barium in the salt and the sodium, potassium or lithium in the zeolite and the occlusion of up to about 25 wt % of the salt within the molecular cavities of the zeolite. This method has the advantage over many methods in that the radionuclides and barium are concentrated in the zeolite so that some of the salt partially purified of cesium, strontium and barium might be recycled back to the electrorefiner. Although this method is effective for purifying the salt, the method requires the removal of the non-occluded surface salt from the zeolite before it can be sent to storage. Furthermore, problems have been encountered in making dense, leach-resistant waste forms directly from the salt-occluded waste product. The use of synthetic naturally occurring minerals to store radioactive ions has also been studied. U.S. Pat. No. 4,808,318 describes the use of a modified phlogopite to recover cesium ions from waste solutions. The modified phlogopite containing the cesium ion is then fixed so that it can be safely stored for a long period of time. U.S. Pat. No. 4,229,317 describes a method whereby radioactive iodine, present as alkali metal iodides or iodates is incorporated into a solid by adding appropriate amounts of alkali metal, alumina and silica to the solution, stirring to form a homogenous mixture, drying the mixture to form a powder and compacting and heating the powder under conditions appropriate to form sodalite, whereby the iodine ion is incorporated within the molecular cage of the sodalite (Na.sub.6 [(SiO.sub.2).sub.6 (AlO.sub.2).sub.6 ]2NaCl). What is still needed is a method of immobilizing mixtures of salts, particularly chloride salts containing radionuclides and other hazardous wastes, so that the highly soluble salts can be safely stored for long periods of time in HLW storage facilities without presenting a hazard to the environment. SUMMARY OF THE INVENTION A method has been found by which mixed waste chloride salts containing radionuclides and other hazardous wastes can be incorporated into a synthetic, naturally-occurring mineral to form a leach resistant compact suitable for long-term storage. Furthermore, the method of the invention is compatible for use with the salt-occluded zeolite prepared as described in previously cited U.S. patent application Ser. No. 744,753. The method of the invention for immobilizing waste chloride salts containing radionuclides and hazardous material for permanent disposal comprises forming a mixture of an effective amount of aluminum oxide, silicon dioxide and sodium hydroxide with respect to the formation of a sodalite, Na.sub.6 [(SiO.sub.2).sub.6 (AlO.sub.2).sub.6 ].y[(A)(X).sub.z ], where y is greater than 0.5 and less than or equal to 2.0, A is an alkali metal or alkaline earth, X is a halide, and z is either 1 or 2, heating the mixture to a temperature sufficient to partially react the mixture to form water and a sodalite intermediate, maintaining the temperature for a period of time sufficient to drive off the water and to form a water-free sodalite intermediate, mixing the water-free sodalite intermediate with from about 5 to 13 wt. percent mixed chloride salts containing radionuclides and hazardous material to form a waste mixture, and heating the mixture to a temperature and for a period of time sufficient to form sodalite, whereby the chloride salt, the radionuclides and the hazardous material are incorporated into the sodalite, thereby immobilizing the waste chloride salt containing radionuclides and hazardous materials. Preferably, the water-free sodalite intermediate is mixed with the salt-occluded zeolite to form the waste mixture, the waste mixture containing from 8 to 13 wt % chloride salt, radionuclides and hazardous material. The advantage of mixing the salt-occluded zeolite is that the radionuclides and hazardous material is much more concentrated in the zeolite than it is in the waste salt alone. It is therefore one object of the invention to provide an effective method for disposing of the waste chloride salt. It is another object of the invention to provide an improved method for stabilizing waste chloride salts containing radionuclides and other hazardous waste materials. It is still another object of the invention to provide an improved method for stabilizing waste chloride salts containing radionuclides and other hazardous waste materials so that they may be safely placed in high-level waste facilities for long periods of time without fear of damage to the environment. It is still another object of the invention to provide an improved matrix material for storing waste chloride salts containing radionuclides such as cesium and strontium and other hazardous wastes such as barium so that they may be safely stored for long periods of time without causing damage to the environment. Finally it is the object of the invention to provide an improved method of stabilizing zeolite-occluded waste chloride salts containing strontium, cesium and barium, so that they may be safely stored for long periods of time without fear of causing damage to the environment. DETAILED DESCRIPTION OF THE INVENTION These and other objects of the invention may be met by first preparing sodalite intermediate by intimately mixing 2 moles of NaOH, 1 mole of Al.sub.2 O.sub.3, and 2 moles of SiO.sub.2, heating the mixture to a temperature of between 250.degree. and 600.degree. C., preferably between 300.degree. and 500.degree. C., for 2 to 20 hours to drive off water and form a water-free sodalite intermediate consisting essentially of a reactive mixture of NaAlO.sub.2, Na.sub.2 SiO.sub.3, Al.sub.2 O.sub.3 and SiO.sub.2. The product is then cooled in a dry atmosphere to about room temperature and then ground to a particle size of between 50 and 500 .mu.m. The powdered sodalite intermediate is then mixed with either waste chloride salt containing radionuclides and hazardous material or salt-occluded zeolite, which been previously ground to a similar size, in amounts such that the mixture contains between 5 and 13 wt % chloride salts. The sodalite intermediate-waste salt mixture is first compacted at 250.degree. to 500.degree. C. and at pressures from about 10 up to 70 MPa or greater for 1 to 8 hours to form a green compact. The compact is then reacted by either maintaining the pressure and raising the temperature to between 700.degree. and 900.degree. C. for 20 to 200 hours or heating the green compact in a closed container to 700.degree. to 900.degree. C. for 20 to 200 hours to react the sodalite intermediate to form sodalite. The product waste form consists of the salt and the radioactive and hazardous components encapsulated in the molecular structure of the sodalite. Preferably once the water-free sodalite intermediate has been formed, it is maintained in a water-free environment to prevent reabsorption of water which may later affect the quality of the final product. As described in the reference patent application, the salt-zeolite product may be prepared by contacting molten waste chloride salt containing the chlorides of cesium, strontium, barium and other radioactive and hazardous waste components with dehydrated zeolite in the sodium, lithium, or potassium form, said zeolite containing molecular cavities, maintaining the contact at 400.degree. to 500.degree. C. for up to 20 hours, a period of time sufficient for the salt to penetrate the zeolite cavities thereby occluding salt within the zeolite and for cesium, strontium and barium in the non-occluded salt to ion-exchange with the sodium, lithium, or potassium in the zeolite. After cooling, the resultant material consists of zeolite with the ion-exchanged cesium, strontium, and barium and occluded salt in the molecular cavities, and with salt adhering to the external surfaces of the zeolite particles. Using the invention, it is not necessary to remove large fractions of the surface salt to make a leach resistant waste form, because the surface and occluded salt are contained in the sodalite molecules. It is desirable to remove as much of the surface salt as possible from the zeolite-salt product to minimize waste volumes. The formation of sodalite appears to proceed in two stages: During the first stage, the sodalite intermediate is formed by the following reaction: EQU 2NaOH+SiO.sub.2 =Na.sub.2 SiO.sub.3 +H.sub.2 O, and EQU 2NaOH+Al.sub.2 O.sub.3 =2NaAlO.sub.2 +H.sub.2 O. Following these reactions to form the intermediate and water, sodalite is formed by, (6-2.alpha.)NaAlO.sub.2 +.alpha.Na2SiO.sub.3 +.alpha.Al.sub.2 O.sub.3 +(6-.alpha.)SiO.sub.2 +(chloride salt or salt-occluded zeolite)=[sodalite] where .alpha. represents the variability in the fraction of the sodium hydroxide reacted with silica or alumina of the first stage reactions. When zeolite is used in the second stage, the zeolite may be transformed into other aluminosilicate compounds, for example sodalite. It is important that the reaction conditions during the first stage be maintained for a period of time sufficient for the water resulting from the reaction to be driven off so that a water-free intermediate results. This prevents water from reacting with the chloride salt in the second stage which would result in corrosive conditions in the reactor. Also, any water present during the second stage may cause the formation of glassy phases or aluminosilicates without molecular cages which do not have the capability for containing chloride salt. Preferably the sodalite intermediate is mixed with the powdered salt-zeolite since the radioactive and hazardous components can be concentrated in this material. Alternatively, the sodalite intermediate may be mixed directly with the waste salt containing the radionuclides and hazardous material. The mixture may contain at least 5 and no more than about 13 wt % of the chloride salts including radioactive and hazardous components. Amounts less than about 5 wt % will result in a waste form with an excessive volume while amounts greater than about 13 wt % will not result in the incorporation of all the salt in the sodalite molecular cages. Preferably, the mixture will contain about 8 to 11 wt % of the salt including the radioactive and hazardous components. Preferably the sodalite-intermediate, the waste salt and/or the salt-occluded zeolite are powdered before mixing to form the waste mixture in order to facilitate the intimate mixing of the components and formation of the green compact. The powder can be formed by any convenient means. A powder size of from about 50 to 500 .mu.m has been found satisfactory. Once the salt-occluded zeolite-intermediate or salt-intermediate mixture has been prepared and well mixed, it can be compacted to form a green pellet. While the conditions for preparation of the green pellet are not critical, heating at 325.degree. C. in a uniaxial press under a pressure of about 70 MPa for about 4 hours was found to prepare a suitable compact. This step should be done in a dry inert atmosphere. The green compact must be heated to a temperature and for a period of time sufficient to form the sodalite. Preferably, the heating takes place in a closed vessel to prevent volatilization of the salts or radionuclides. The temperature may vary from about 700.degree. C. to about 1000.degree. C., preferably from about 700.degree. C. to 900.degree. C. Sodalite formation required temperatures greater than 700.degree. C. while decomposition begins at temperatures greater than 1000.degree. C. A heating period of 20 to 200 hours was found sufficient for sodalite formation. This step should be done in a dry, inert atmosphere. Alternatively to forming a compact and then heating the compact to form the sodalite, the sodalite can be formed directly from the salt-intermediate mixture by placing this mixture in a hot press and heating under a pressure of about 70 MPa to a temperature of 700.degree. to 800.degree. C. and for a period of time up to 200 hours. |
claims | 1. A nuclear waste hazardous material storage bank, comprising:a wellbore extending into the Earth and comprising an entry at least proximate a terranean surface, the wellbore comprising a substantially vertical portion, a transition portion, and a substantially horizontal portion;a storage area coupled to the substantially horizontal portion of the well bore, the storage area within or below an impermeable, hydrocarbon bearing shale formation, the storage area vertically isolated, by the hydrocarbon bearing shale formation, from a subterranean zone that comprises mobile water, the shale formation comprising a diffusion barrier to a radioactive gas isotope of the nuclear waste hazardous material;a storage container positioned in the storage area, the storage container sized to fit from the wellbore entry through the substantially vertical, the transition, and the substantially horizontal portions of the wellbore, and into the storage area, the storage container comprising an inner cavity that encloses a nuclear waste hazardous material; anda seal positioned in the wellbore, the seal isolating the storage area of the wellbore from the entry of the wellbore. 2. The nuclear waste hazardous material storage bank of claim 1, wherein the storage area is formed below the hydrocarbon bearing shale formation and is vertically isolated from the subterranean zone that comprises mobile water by the hydrocarbon bearing shale formation. 3. The nuclear waste hazardous material storage bank of claim 1, wherein the storage area is formed within the hydrocarbon bearing shale formation, and is vertically isolated from the subterranean zone that comprises mobile water by at least a portion of the hydrocarbon bearing shale formation. 4. The nuclear waste hazardous material storage bank of claim 1, wherein the hydrocarbon bearing shale formation comprises a permeability of less than about 0.001 millidarcys. 5. The nuclear waste hazardous material storage bank of claim 1, wherein the hydrocarbon bearing shale formation comprises a brittleness of less than about 10, where brittleness comprises a ratio of compressive stress of the hydrocarbon bearing shale formation to tensile strength of the shale formation. 6. The nuclear waste hazardous material storage bank of claim 1, wherein the hydrocarbon bearing shale formation comprises a thickness proximate the storage area of at least about 100 feet. 7. The nuclear waste hazardous material storage bank of claim 1, wherein the hydrocarbon bearing shale formation comprises a thickness proximate the storage area that inhibits diffusion of the nuclear waste hazardous material that escapes the storage container through the hydrocarbon bearing shale formation for an amount of time that is a multiple of a half-life of a radioactive component of the nuclear waste hazardous material. 8. The nuclear waste hazardous material storage bank of claim 1, wherein the hydrocarbon bearing shale formation comprises about 20 to 30% weight of clay or about 20 to 30% weight of organic matter. 9. The nuclear waste hazardous material storage bank of claim 1, wherein the nuclear waste hazardous material comprises spent nuclear fuel. 10. The nuclear waste hazardous material storage bank of claim 1, further comprising at least one casing assembly that extends from at or proximate the terranean surface, through the wellbore, and into the storage area. 11. The nuclear waste hazardous material storage bank of claim 1, wherein the storage container comprises a connecting portion configured to couple to at least one of a downhole tool string or another storage container. 12. A method for storing nuclear waste hazardous material, comprising:moving a storage container through an entry of a wellbore that extends into a terranean surface, the entry at least proximate the terranean surface, the storage container comprising an inner cavity that encloses nuclear waste hazardous material;moving the storage container through the wellbore that comprises a substantially vertical portion, a transition portion, and a substantially horizontal portion, the storage container sized to fit from the wellbore entry through the substantially vertical, the transition, and the substantially horizontal portions of the wellbore;moving the storage container into a storage area that is coupled to the substantially horizontal portion of the well bore, the storage area located within or below an impermeable, hydrocarbon-bearing shale formation and vertically isolated, by the hydrocarbon-bearing shale formation, from a subterranean zone that comprises mobile water, the shale formation comprising a diffusion barrier to a radioactive gas isotope of the nuclear waste hazardous material; andforming a seal in the wellbore that isolates the storage portion of the wellbore from the entry of the wellbore. 13. The method of claim 12, wherein the storage area is formed below the hydrocarbon-bearing shale formation and is vertically isolated from the subterranean zone that comprises mobile water by the hydrocarbon-bearing shale formation. 14. The method of claim 12, wherein the storage area is formed within the hydrocarbon-bearing shale formation, and is vertically isolated from the subterranean zone that comprises mobile water by at least a portion of the hydrocarbon-bearing shale formation. 15. The method of claim 12, wherein hydrocarbon-bearing the shale formation comprises geological properties comprising two or more of:a permeability of less than about 0.001 millidarcys;a brittleness of less than about 10, where brittleness comprises a ratio of compressive stress of the hydrocarbon-bearing shale formation to tensile strength of the hydrocarbon-bearing shale formation;a thickness proximate the storage area of at least about 100 feet; orabout 20 to 30% weight of organic material or about 20 to 30% weight of clay. 16. The method of claim 12, wherein the nuclear waste hazardous material comprises spent nuclear fuel. 17. The method of claim 12, wherein the wellbore further comprises at least one casing that extends from at or proximate the terranean surface, through the wellbore, and into the storage area. 18. The method of claim 12, further comprising:prior to moving the storage container through the entry of the wellbore that extends into the terranean surface, forming the wellbore from the terranean surface to the hydrocarbon-bearing shale formation. 19. The method of claim 18, further comprising installing a casing in the wellbore that extends from at or proximate the terranean surface, through the wellbore, and into the storage area. 20. The method of claim 19, further comprising cementing the casing to the wellbore. 21. The method of claim 18, further comprising, subsequent to forming the wellbore, producing hydrocarbon fluid from the hydrocarbon-bearing shale formation, through the wellbore, and to the terranean surface. 22. The method of claim 12, further comprising:removing the seal from the wellbore; andretrieving the storage container from the storage area to the terranean surface. 23. The method of claim 12, further comprising:monitoring at least one variable associated with the storage container from a sensor positioned proximate the storage area; andrecording the monitored variable at the terranean surface. 24. The method of claim 23, wherein the monitored variable comprises at least one of radiation level, temperature, pressure, presence of oxygen, presence of water vapor, presence of liquid water, acidity, or seismic activity. 25. The method of claim 23, further comprising, based on the monitored variable exceeding a threshold value:removing the seal from the wellbore; andretrieving the storage container from the storage area to the terranean surface. 26. A spent nuclear fuel storage system, comprising:a directional wellbore formed from a terranean surface, through a first subterranean layer, and into a second subterranean layer deeper than the first subterranean layer, the first subterranean layer comprising a rock formation that includes a source of mobile water, the second subterranean layer comprising an impervious, hydrocarbon bearing shale formation, the shale formation fluidly isolating a portion of the directional wellbore formed within the shale formation from the first subterranean layer, the shale formation comprising a diffusion barrier to a radioactive gas isotope of the nuclear waste hazardous material;a container configured to be moved through the directional wellbore into the portion of the directional wellbore formed within the shale formation, the container comprising a volume enclosed by a housing configured to store a plurality of spent nuclear fuel pellets; anda plug set in the directional wellbore between the portion of the directional wellbore formed within the shale formation and the terranean surface. 27. The spent nuclear fuel storage system of claim 26, further comprising a monitoring system, comprising a monitoring control system communicably coupled to one or more systems positioned proximate the container. 28. The spent nuclear fuel storage system of claim 26, further comprising a tubular liner constructed in the directional wellbore and sealed against a wall of the directional wellbore. 29. The nuclear waste hazardous material storage bank of claim 1, wherein at least a part of the substantially horizontal portion of the wellbore defines a volume that comprises the storage area, and the part of the substantially horizontal portion of the wellbore is formed within the hydrocarbon-bearing shale formation. 30. The method of claim 12, further comprising repairing perforations in a production casing in the wellbore prior to moving the storage container into the storage area. 31. The nuclear waste hazardous material storage bank of claim 1, wherein the hydrocarbon-bearing shale formation is at a true vertical depth of between 3000 and 12,000 feet. 32. The nuclear waste hazardous material storage bank of claim 1, wherein the radioactive gas isotope comprises tritium. 33. The nuclear waste hazardous material storage bank of claim 1, wherein the diffusion barrier comprises a diffusion time of the radioactive gas isotope of the nuclear waste hazardous material of a multiple of a half-life of the radioactive gas. 34. The nuclear waste hazardous material storage bank of claim 33, wherein the multiple is between thirty and fifty times the half-life of the radioactive gas isotope. |
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description | The present invention relates to components in an ion implanter that may see incidence of the ion beam, such as a beam dump or a beam stop. Ion implanters are used in the manufacture of semiconductor devices and other materials. In such ion implanters, semiconductor wafers or other substrates are modified by implanting atoms of a desired species into the body of the wafer, for example to form regions of varying conductivity. Ion implanters are well known and generally conform to a common design as follows. An ion source generally comprises an arc chamber in which a hot plasma is generated. The plasma will contain ions of a desired species to be implanted. An extraction lens assembly produces an electric field that extracts ions from the ion source and forms a mixed beam of ions. Only ions of a particular species are usually required for implantation in a wafer or other substrate, for example a particular dopant for implantation in a semiconductor wafer. The required ions are selected from the mixed ion beam that emerges from the ion source by using a mass analysing magnet in association with a mass-resolving slit. By setting appropriate operational parameters on the mass-analysing magnet and the ion optics associated therewith, an ion beam containing almost exclusively the required ion species emerges from the mass-resolving slit. The ions travel along a flight tube as they pass through the mass-analysing magnet. The ion beam is transported along a beam line to a process chamber where the ion beam is incident on a substrate held in place in the ion beam path by a substrate holder. The substrate may be a semiconductor wafer. The various parts of the ion implanter are operated under the management of a controller, typically a suitably trained person, a programmed computer, or the like. A more detailed description of an ion implanter of this general type can be found in U.S. Pat. No. 4,754,200. Ions may strike some components within the ion implanter relatively frequently (other than the substrate to be implanted). For example, ions with a large mass-to-charge ratio will not be deflected sufficiently by the mass-resolving magnet to pass through the mass-resolving slit. As a result, a beam dump may be provided to adsorb such ions. These ions striking the beam dump may cause sputtering of material. Care must be taken though, as material sputtered from the beam dump may become entrained within the ion beam and so contaminate the substrate. In addition, there are times when the ion beam may be dumped into the beam dump on purpose. For example, instability in the ion beam may require that implantation of a wafer be stopped as quickly as possible. One way of achieving this is to switch off the mass-analysing magnet. With the magnet switched off, the ions merely follow a straight path rather than the usual curved path through the flight tube. The beam dump is positioned to absorb the ion beam when it is dumped in this way. Such a beam strike of the whole beam is likely to sputter more material. Although the material can no longer become entrained within the ion beam, there remains a problem in that the beam dump often has line of sight to the substrate. Consequently, material sputtered from the beam dump may still contaminate the substrate. A further example of a component that frequently sees beam strike is the beam stop that resides downstream of the substrate. The ion beam may strike the beam stop when the substrate is moved away from the ion beam path, e.g. during mechanical scanning of the wafer during implants with a spot beam. Unwanted material that has been sputtered from components such as a beam dump may travel to the substrate and subsequently the material may strike the substrate causing contamination or even damage to the devices being formed on the substrate. Moreover, sputtered material may adhere to another surface within the ion implanter. Surfaces adjacent to the ion beam are the most prone to receiving such deposits. As the amount of material deposited accumulates, the chances of the deposits delaminating to form flakes or particles increases. These flakes or particles frequently detach from their host surface and may become entrained in the ion beam. As a result, the flakes or particles contain sputtered material that still ultimately reaches the substrate. Against this background, and from a first aspect, the present invention resides in a method of operating an ion implanter comprising: producing an ion beam; receiving ions from the ion beam in a component having an entrance opening and an internal surface for absorbing ions that have passed through the entrance opening; providing an electrical bias on the internal surface so as to decelerate the ions prior to them striking the internal surface. From a second aspect the present invention resides in an ion implanter comprising power supply apparatus and an ion-receiving component with an entrance opening providing line of sight to an internal surface. The component is arranged to receive ions from an ion beam through the entrance opening such that ions strike the internal surface. The power supply apparatus is arranged to provide an electrical bias to the surface to decelerate the ions prior to their striking the surface. Biasing the surface in this way is advantageous in that it reduces the energy of the ions before they strike the internal surface. Thus, with their energy reduced, the ions will pose less of a problem in sputtering material from the surface. Preferably, the power supply apparatus is arranged to bias the internal surface to be at substantially the same potential as the ion beam. The component may further comprise an array of electrodes disposed between the surface and entrance opening. This allows further electrical control. For example, the array of electrodes may comprise one or more upstream electrodes disposed adjacent the opening. The one or more upstream electrodes may be electrically biased by the power supply to be at substantially the same potential as the ion beam. This is beneficial in that it stops ions within the ion beam travelling past the beam dump, but not travelling into the beam dump, from seeing the potential of the surface. Thus, such ions are not disturbed in their flight by the repulsive electrical field exerted by the surface. In addition, or as an alternative, the array of electrodes may further comprise one or more downstream electrodes positioned adjacent the surface. The one or more downstream electrodes may be electrically biased to repel electrons liberated from the surface. This suppresses these electrons that may otherwise neutralise ions in the beam. The ion receiving component may be a beamstop or it may be part of a flight tube of a mass resolving analyser. The ion receiving component may be used elsewhere in an ion implanter, preferably in positions where it may receive ions, either the ion beam itself or ions that are lost from the ion beam. Other preferred, but optional features, are to be found in the appended claims. In order to provide a context for the present invention, an exemplary application is shown in FIG. 1, although it will be appreciated this is merely an example of an application of the present invention and is in no way limiting. FIG. 1 shows an ion implanter 10 for implanting ions in semiconductor wafers 12 that may be used in accordance with the present invention. The ion implanter 10 comprises a vacuum chamber 15 pumped through valve 24. Ions are generated by ion source 14 and are extracted by an extraction lens assembly 26 to form an ion beam 34. In this embodiment this ion beam 34 is steered and shaped through the ion implanter 10 such that the ion beam 34 passes through a mass analysis stage 30. Ions of a desired mass are selected to pass through a mass resolving slit 32 and then conveyed onward along an ion beam path 34 towards the semiconductor wafer 12. In this embodiment, the ions are decelerated before reaching the semiconductor wafer 12 by deceleration lens assembly 48 and pass through a plasma flood system 49 that acts to neutralise the ion beam 34. Ions formed within the ion source 14 are extracted through an exit aperture 28 using a negatively-biased (relative to ground) extraction electrode 26. A potential difference is created between the ion source 14 and the following mass analysis stage 30 by a power supply 21 such that the extracted ions are accelerated. The ion source 14 and mass analysis stage 30 are electrically isolated from each other by an insulator (not shown). The mixture of extracted ions are then passed through the mass analysis stage 30 so that the mixture passes around a curved path through a flight tube 46 under the influence of a magnetic field. A beam dump 100 resides within the flight tube 46. The radius of curvature traveled by any ion is determined by its mass, charge state and energy. The magnetic field is controlled so that, for a set beam energy, only those ions with a desired mass-to-charge ratio energy exit along a path coincident with the mass-resolving slit 32. The ion beam 34 is then transported to the wafer 12 to be implanted (or other substrate) or to a beam stop 38 when there is no wafer 12 in the target position. Before arriving at the wafer 12 or beamstop 38, the ions may be decelerated using a deceleration lens assembly like that shown at 48 positioned between the mass analysis stage 30 and upstream of the wafer 12. The deceleration lens assembly 48 is followed by a plasma flood system 49 that operates to produce a flood of electrons that are available to the semiconductor wafer 12 to neutralise the effect of the incident positive ions. The semiconductor wafer 12 is mounted on a wafer holder 36, wafers 12 being successively transferred to and from the wafer holder 36 for serial implantation. Alternatively, parallel processing may be used where many wafers 12 are positioned on a carousel 36 that rotates to present the wafers 12 to the incident ion beam 34 in turn. A controller is shown at 50 that comprises a suitably programmed computer. The controller 50 is provided with software for managing operation of the ion implanter 10. A first embodiment of a ion implanter component according to the present invention is shown in FIG. 2. The component shown is a beam dump 60 that may be placed at various location within an ion implanter, such as the one shown in FIG. 1, to receive the ion beam. For example, the beam dump 60 may be used as a beamstop 38 positioned downstream of the wafer 12, so as to receive the ion beam 34 when the wafer 12 is not in the implant position. As another example, the beam dump 60 may be used in a flight tube 46 so as to receive the ion beam 34 when the magnet of the mass analyser 30 is switched off. Also, such a beam dump 60 may be used to receive ions that do not follow the ion beam path 34 through the flight tube 46, i.e. to receive ions not having the desired mass to charge ratio. The beam dump has a generally box-like shape defined by a top 61, a bottom 62, a back wall 63, a front wall 64 and a pair of end walls 65 (only one of which is visible in FIG. 2). The front wall 64 is provided with a central aperture 66 that penetrates through the front wall 64. The beam dump 60 is positioned such that the aperture 66 faces the ion beam 34, so as to receive the ion beam 34 as shown in FIG. 2. The ion beam 34 passes through the aperture 66 and passes a pair of opposed suppression electrodes 67 positioned just beyond the aperture 66. The purpose of the suppression electrodes will be described below. Once past the suppression electrodes 67, the ion beam 34 enters and strikes a cup 68 comprising a base 69 and a cylindrical wall 70. The cup 38 need not be cylindrical, but could be other shapes. The cup 38 is electrically biased, as will now be described. The ion beam 34 has a beam energy equal to the potential set on the ion source 14, e.g. if the ion source 14 is set at +10 kV, ions within the ion beam will typically have an energy of 10 keV. Such high-energy beams 34 are commonly used within ion implanters 10 to reduce the problems of space charge blow-up. Ions striking the cup 68 of the beam dump 60 causes sputtering of material and the problem of material being sputtered from the beam dump 100 worsens the greater the energy of the incident ions. This problem is mitigated by using a power supply unit 71 to place a potential on the cup 68 that decelerates ions in the ion beam 34 before they strike the cup 68. The potential set on the cup 68 is matched to the beam energy and so chosen to be at or preferably just below the potential of the ion source 14. For example, the cup 68 may be biased to be +9.9 kV. In this way, the incoming ions are decelerated to near-zero energy prior to striking the cup 68. Thus, the problem of material being sputtered from the cup 68 is lessened. An alternative to using a power supply unit 71 to provide the decelerating potential is to connect electrically the cup 68 to the ion source 14, such that both are at the same potential. Setting the cup 68 to be at the same potential as the ion source 14 may cause some ions to be reflected by the cup 68, hence a slightly lower potential is preferred. As shown in FIG. 2, the decelerating ion beam 34 has an ever-increasing tendency to blow-up due to space charge effects. A power supply unit 72 is used to set a potential on the suppression electrodes 67. Power supply units 71 and 72 may be combined if desired. The suppression electrodes 114 are set at a high negative potential, for example −5 kV. This is to suppress electron travel in either direction. In particular, the suppression electrodes 114 suppress any electrons liberated from the cup 68 from travelling back out of the beam dump 60. Such an electron beam may otherwise cause damage within the ion implanter 10. For example, the electron beam may cause heating of any part it impacts and this can be extreme enough to cause melting. Obviously, the potential for any electron beam striking the wafer 12 to cause serious damage is considerable. Electron impact may also cause x-ray emission. FIG. 3 shows a representation of the mass analyser 30 of FIG. 1, along with the path 34 of ions through a flight tube 46 defined by the mass analyser 30. The solid line 34 shows the path of ions having the desired mass-to-charge ratio and describes a smooth quarter-turn through the mass analyser 30. While the beam dump 60 of FIG. 2 may be used in this flight tube 46, FIG. 3 shows an alternative embodiment of a beam dump 100. The beam dump 100 is provided for ions having a greater mass-to-charge ratio than desired, and for instances when the ion beam 34 is dumped. Ions having a greater mass-to-charge ratio than desired may strike the beam dump 100 as shown at 101-104. The path that the ion beam follows when the magnet of the mass analyser 30 is switched off is shown at 105. Ions having a lesser mass-to-charge ratio than desired will turn inwardly from the ion beam path 34. Although not shown, a further beam dump 60, 100 may be provided on the inner radius of the path 34 to receive such lighter ions. Ions that strike the beam dump 100 may sputter material. Typically, beam dump 100 will be made from graphite and so there is a danger that graphite will become entrained in the ion beam 34 as it passes through the mass analyser 30. This entrained material may be deposited on nearby parts, causing deposited coatings that can then flake off, generating particulates. These particulates can then be transported to the wafer 12, causing contamination. FIGS. 4 to 6 show beam dump 100 in greater detail. The beam bump 100 is broadly box like and has a dog-legged shape 106. The beam dump 100 comprises side walls 107 and 108, a base 109 and two back walls 110 and 111. Thus, the beam dump 100 has an open front face entrance opening) 112 to allow entry of ions from the ion beam 34. A graphite dump plate 113 is attached to the back walls 110 and 111 by any convenient means, e.g. screws, bolts, etc. The dump plate 113 has the same dog-leg shape to conform to the shape of the back walls 110-111. Sitting in front of the dump plate 113 within the beam dump 100 are two sets of electrodes 114 and 115. The electrodes may be made from tungsten, or other materials such a graphite, stainless steel, etc. Each set of electrodes 114-115 comprises four identical generally planar electrodes 114a-d and 115a-d that are arranged one above another. Each electrode 114a-d and 115a-d extends from one side wall 107 to the other side wall 108, and has the common dog-leg shape. The electrodes may be fixed in place in any convenient manner. Electrodes from each set are paired with one another, such that electrode 114a resides adjacent to electrode 115a, and so on. As the electrodes 114-115 extend from near top to near bottom of the beam dump 100, they effectively present a grill to ions entering the beam dump 100, i.e., their front edges face the entrance opening 112. How the electrodes 114-115 and the dump plate 113 are advantageously biased will now be described. For the sake of clarity, the power supplies for and the electrical connections to the electrodes 114-115 and dump plate 113 are not shown in FIGS. 4 to 6. Nonetheless, the person skilled in the art will readily identify many different ways of arranging the electrical connections and supplies. As described above, the ion beam 34 has a beam energy equal to the potential set on the ion source 14, e.g. 10 keV The potential set on the dump plate 113 is matched to the beam energy and so chosen to be at or preferably just below the potential of the ion source 14, e.g. +9.9 kV. Hence, the incoming ions are decelerated to near-zero energy prior to striking the dump plate 113 and so the problem of material being sputtered from the dump plate 113 is lessened. To ensure that the potential of the dump plate 113 is not seen by ions before they enter the beam dump 100, the potential set on the front set of electrodes 115 is the same as the surrounding beamline. This may be achieved most easily by linking the potential of the front set of electrodes 115 to that of the flight tube 46 or the surrounding parts. Ensuring that the potential of the dump plate 113 is not seen by ions before they enter the beam dump 100 is important from the point of view of ions within the ion beam 34 that have the desired mass-to-charge ratio, as they should pass through the mass analyser 30 undisturbed by stray electric fields. The back set of electrodes 114 are used to suppress electron travel and so is set at a high negative potential (relative to the front set of electrodes 115), for example −2 kV. In particular, the back set of electrodes 114 suppresses any electrons liberated from the dump plate 113 from travelling back out of the beam dump 100. As described above, such an electron beam may otherwise cause damage within the ion implanter 10 or to the wafer 12. As will be appreciated by the person skilled in the art, variations may be made to the above embodiment without departing from the scope of the invention defined by the claims. For example, it will be realised that the terms front, back, sides and base used above are merely relative and that the beam dump 60, 100 may be used in any orientation. As a result, the terms may need to be changed according to the particular orientation of the beam dump 60, 100 chosen. Various features of the beam dumps 60, 100 may be interchanged between the two designs. For example, one or more screening electrodes may be used in the beam dump 60 of FIG. 2: an array of electrodes akin to the front set of electrodes 115 of FIGS. 4 to 6 may be used, or either a single such electrode or pair of electrodes may be used. The screening electrodes should have a negative potential to suppress electron travel. As another example, the cup 68 of the beam dump 60 of FIG. 2 may be used in place of the dump plate 113 of FIGS. 4 to 6. Such a cup 68 may be advantageous as it reduces the risk of the ion beam 34 missing the dump plate 113 (remembering that the ion beam 34 is prone to blow-up as it approaches the dump plate 113, in the manner shown in FIG. 2). FIGS. 1 and 3 show an ion implanter 10 with a single beam dump 100 provided in the flight tube 46. However, two or more beam dumps 100 may be provided. For example, a series of beam dumps 60, 100 may be provided around the outer radius of the ion beam path 34 through the flight tube 46. The beam dumps 60, 100 may be progressively angled to follow approximately the ion beam path 34 through the flight tube 46. Although, beam dumps 60, 100 have been described in use as a beamstop 38 and within a flight tube 46, the beam dumps 60, 100 may be used in any position within an ion implanter 10 that may receive ions from the ion beam 34. The dog-leg design of the beam dump 100 of FIGS. 4 to 6 need not be used: as well as the linear beam dump 60, other shapes may be used. |
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051788225 | abstract | In combination with a steam generator having a plurality of generator tube support plates, each generator tube support plate having a plurality of openings and a plurality of generator tubes, each tube passing through aligned openings in the support plates, a corrosion monitoring system is provided including a mockup probe, comprising a probe tube support plate having an upper side and a lower side and having substantially the same thickness and being constructed of substantially the same material as the generator tube support plates, having at least one opening of substantially the same size and shape as the openings of the generator tube support plates; at least one probe tube having an upper end and a lower end and having substantially the same diameter as the generator tubes and being constructed of substantially the same material as the generator tubes, the probe tube passing through the opening of the probe tube support plate; and wherein the mockup probe is adapted such that it may be inserted and sealed within the steam generator during chemical cleaning operations. The mockup probe is preferably used in combination with electronic linear polarization and zero resistance ammetry probes, which electronically measure corrosion during cleaning operations. |
description | This invention relates generally to radiation apparatuses and methods, and in particular to multileaf collimators and methods of adjusting radiation beams useful in radiotherapy and other industries. Multileaf collimators (MLCs) are widely used in radiotherapy machines to support various treatments including intensity-modulated radiation therapy (IMRT) and arc therapy, etc. Conventional multileaf collimators include a single level of a plurality of beam blocking leaves arranged in two opposing banks or arrays. Each leaf in a bank is longitudinally movable relative to a leaf in the opposing bank. In operation each of the individual leaves is positioned to block a portion of a radiation beam passing through the volume occupied by the leaf. The combined positioning of all leaves defines one or many apertures through which the unblocked radiation beam passes, and the aperture(s) define(s) the shape of the radiation beam directed to a treatment field at an isocenter. To mitigate radiation leakage in single level MLCs, various leaf designs are developed including “tongue in groove” designs in which steps, waves or similar geometries are provided on the leaf sides so that leaf materials mutually overlap between leaves as viewed from a radiation source. While a tongue in groove design may reduce leakage between leaf sides, it unfortunately leads to undesirable underdose effects when MLC treatment fields are combined. Some conventional MLCs are used in combination with one or two pairs of collimation jaws to reduce leakage between abutted leaf ends. One issue associated with the combination of a MLC with collimation jaws is the increased bulk of a radiation system and the resulting reduced clearance between the patient and moving equipment. It is desirable to provide MLCs that can shape beams with high resolution so that the shaped beam conforms to a target volume as close as possible. In general a MLC would provide for higher beam shaping resolution if the beam blocking leaves could be thinner. However, reducing the width of leaves to improve MLC resolution has limitations and imposes challenges to MLC construction and operation. For MLCs using screw leaf drive systems for example, long slender drive screws may be susceptible to column buckling in a way that scales dramatically worse with smaller screw diameters. Motors with a smaller diameter may also be required. This invention provides for multi level MLCs and methods of shaping beams that can significantly reduce various leakage effects and improve beam shaping resolution. In some embodiments, a multilevel MLC comprises a first set and a second set of a plurality of pairs of beam blocking leaves arranged adjacent to one another. Leaves of each pair in the first set are disposed in an opposed relationship and longitudinally movable relative to each other in a first direction. Leaves of each pair in the second set are disposed in an opposed relationship and longitudinally movable relative to each other in a second direction generally parallel to the first direction. The first and second sets of pairs of leaves are disposed in different planes. In some embodiments, each of the first and second sets includes a first section of a plurality of pairs of leaves having a first cross section and a second section of a plurality of pairs of leaves having a second cross section different from the first cross section. In some embodiments, the first cross section of the leaves in the first section of the first set is different from the first cross section of the leaves in the first section of the second set. In some embodiments, the leaves in the first and second sets substantially focus on a single converging point. The leaves may have a trapezoidal cross section and generally flat side surfaces. Each leaf in the first set may be offset from a leaf in the second set in a direction generally traverse to the first and second directions. The leaves in the first and second sets may be supported by one or more movable carriages. In some embodiments, each leaf in the first set is offset from a leaf in the second set by substantially half the leaf in a direction generally traverse to the first and second directions. The leaves in the first set may have a substantially same first cross section and the leaves in the second set may have a substantially same second cross section. In one aspect a method of shaping radiation beams using a multi level MLC is provided. The multi level MLC comprises first and second sets of a plurality of beam blocking leaves disposed in first and second planes. Leaves in each of the first and second sets are arranged in two opposing arrays forming a plurality of pairs of leaves in the first and second sets respectively. Leaves of each pair are arranged in an opposed relationship and longitudinally movable relative each other, and the longitudinal moving directions are substantially parallel generally traverse to a beam direction. The leaves in the first and second sets are moved to block a selected portion of a radiation beam. In moving the leaves to produce treatment fields, generally, at least a portion of at least one leaf in an array of the first set overlaps at least a portion of at least one leaf in an opposing array of the second set in the beam direction. In some embodiments, the at least one leaf in the first set can come in contact with a leaf in an opposing array in the first set. In some embodiments, the at least one leaf in the first set can come in contact with a leaf in an opposing array in the first set at a first position, and the at least one leaf in the second set can come in contact with a leaf in an opposing array of the second set at a second position that is offset from the first position in the leaf moving directions. Various embodiments of multi level MLCs are described. It is to be understood that the invention is not limited to the particular embodiments described as such and may, of course, vary. An aspect described in conjunction with a particular embodiment is not necessarily limited to that embodiment and can be practiced in any other embodiments. For instance, while various embodiments are described in connection with X-ray radiotherapy machines, it will be appreciated that the invention can also be practiced in other electromagnetic apparatuses and modalities. It is also to be understood that the terminology used herein is for the purpose of describing particular embodiments only, and is not intended to be limiting since the scope of the invention will be defined by the appended claims, along with the full scope of equivalents to which such claims are entitled. In addition, various embodiments are described with reference to the figures. It should be noted that the figures are not drawn to scale, and are only intended to facilitate the description of specific embodiments. They are not intended as an exhaustive description or as a limitation on the scope of the invention. Various relative terms such as “upper,” “above,” “top,” “over,” “on,” “below,” “under,” “bottom,” “higher,” “lower” or similar terms may be used herein for convenience in describing relative positions, directions, or spatial relationships in conjunction with the drawings. For example, the term “level” or “upper or lower level” may be used for ease of describing some embodiments when a radiation source is on the top of an isocenter and a multi level MLC is positioned therebetween. The use of the relative terms should not be construed as to imply a necessary positioning, orientation, or direction of the structures or portions thereof in manufacturing or use, and to limit the scope of the invention. As used in the description and appended claims, the singular forms of “a,” “an,” and “the” include plural references unless the context clearly dictates otherwise. For example, reference to “a direction” includes the opposite direction of the direction and a plurality of directions that are parallel to the direction. A direction includes both linear and arc trajectories. As used herein the term “support body” may include a single support body member or a support body assembly comprised of a plurality of body members. The term “plane” as used in the plane of beam blocking leaves include both planar and curved or cylindrical planes. In general, the present invention provides a multi level MLC that includes two or more sets of beam blocking leaves in two or more different levels or planes. The two or more sets of leaves may be arranged stacked one above the other and parallel so that all leaves may travel in a substantially same direction. The two or more sets of leaves may also be arranged offset such that each leaf in a set may be offset from a leaf in a different set in a direction generally traverse to the leaf travel direction. FIG. 1 is a simplified illustration of a radiation system 100 that includes an exemplary multi level MLC in accordance with some embodiments of the invention. The radiation system 100 includes a radiation source 102 that is configured to produce beams 103 such as of photons, electrons, protons, or other types of radiation. For example, in X-ray radiotherapy the radiation source 102 may include a target which can produce X-ray radiation when impinged by energetic electron beams. The radiation system 100 may include beam shaping components such as a primary collimator 104 and optionally a secondary collimator 106 to generally limit the extent of the beam as it travels away from the radiation source 102 toward an isocenter 108. A multi level MLC 110 can be disposed between the radiation source 102 and the isocenter 108 to further adjust the shape and/or intensity of the beam 103 projected toward the isocenter 108. The MLC 110 and optionally a secondary collimator 106 may rotate about an axis through the source 102 and the isocenter 108, facilitated by bearing 105. The radiation source 102, primary collimator 104, bearing 105, secondary collimator 106, and MLC 110 may be enclosed in or attached to a structure such as a gantry, which may rotate about an axis such as a horizontal axis 109 through the isocenter 108. Thus in some embodiments, the radiation system 100 can deliver a treatment beam to a target in the isocenter plane 108 from various angles, and the shape and/or intensity of the beam can be dynamically adjusted by the MLC 110 as the beam angle is swept or stepped around the target. The radiation system 100 may also include various other components which are not shown in FIG. 1 in order to simplify the description of the invention. For example, the radiation system 100 may include a flattening filter for providing uniform dose distribution, an ion chamber for monitoring the parameters of a beam, and a field light system for simulation of a treatment field, etc. The radiation system 100 may also optionally include one or two pairs of collimation jaws movable in x- and/or y-directions (lower jaws, upper jaws) to provide for rectangular shaping of beams. In some embodiments, the radiation system 100 may include one of the collimation jaw pairs in conjunction with a multi level MLC of the invention. In some embodiments, the radiation system 100 does not require collimation jaws; the inclusion of a multi level MLC of the invention may effectively replace both the upper and lower jaws. As will be described in greater detail below, the design and control of the multi level MLC of the invention can significantly reduce various leakage effects, thus additional collimation jaws would not be required. Replacement of conventional collimation jaws would be an advantage as it reduces the bulk of a radiation system and improves the clearance between the patient and moving equipment. FIG. 2 is a cross-sectional view of an exemplary multi level MLC 210 in accordance with some embodiments. To simplify description, two sets of beam blocking leaves at two different levels or planes are shown in FIG. 2. It will be appreciated that three or more sets of leaves can be arranged at three or more different levels. As shown, the two or more sets 220, 230 can be arranged stacked and parallel. In each set, a plurality of leaves may be arranged in two banks or arrays forming a plurality of pairs of opposing leaves. Each leaf of a pair in a bank can be longitudinally movable relative to the other leaf of the pair in the opposing bank. In some embodiments, the two or more sets 220, 230 can be arranged such that the leaves at different levels may travel in a same direction. For example, the two or more sets 220, 230 may be arranged such that all the leaves in the MLC 210 can travel in e.g. the x-direction generally traverse to the beam direction when in use. The leaves of the MLC can be supported by a support body 212 which may include such as frames, boxes, carriages or other support structures. In some embodiments all the MLC leaves in different sets 220, 230 can be supported by a single carriage (unicarriage). The single carriage, supporting all the MLC leaves, can be driven such as by a powered actuating mechanism in the MLC leaf travel direction. In some embodiments, the MLC 210 may include two carriages each supports a portion of the MLC leaves or each supports a level of leaves. FIG. 3 illustrates an exemplary two level MLC 310 including two carriages 312, 314. One carriage 312 may support half the MLC leaves on a same side of all levels, and the other carriage 314 supports the other half on the opposing side. The two carriages 312, 314, each supporting half the MLC leaves, can be independently moved by powered actuating mechanism 330 along the MLC leaf travel direction. The carriages 312, 314 may travel on guide rails 316. Numerous arrangements and types of guide rails and powered actuators could be used to support and move carriages. The use of one or more carriages may provide advantages in that individual leaves and their travel can be shorter, and therefore have better tolerance control, less cost, less weight, and can fit in a smaller cover or similar structures. Combined speed of leaves and carriages can be a treatment planning advantage. In some embodiments, the multi level MLC of the invention does not require a movable carriage or carriages (carriageless). As shown in FIG. 3, each of the MLC leaves 318 can be independently moved by an associated drive motor 320. The drive motors 320 can be secured to the support body such as a carriage or carriages 312, 314 and are coupled to position feedback devices, a computer and motion control (not shown). In operation the drive motors 320 receive signals from the computer and motion control and move to position individual leaves 318 relative to the beam direction based on a treatment plan. The positioning of a leaf operates to block or adjust the radiation beam which is passing through the volume occupied by the leaf. The combined positioning of all leaves may define one or more aperture(s) through which an unblocked radiation beam passes, and the aperture(s) may define the shape of the radiation beam projected to a target which may be located in the isocenter plane. Returning to FIG. 1, the shape of the radiation beam 103 projected on the isocenter plane 108 has a step or strip resolution at the beam boundary 112. The step resolution is a function of the width of individual leaves of the MLC 110 and the position of the leaves relative to the isocenter 108 and the radiation source 102 from which the beam is emitted and diverged. In general, the step resolution would be higher if the leaves of the MLC 110 were thinner. Higher step resolution can also be provided by positioning the MLC 110 closer to the isocenter 108. In the description of the MLC definition and various radiation leakage effects, the terms of “length,” “width,” “height,” “side,” and “end” of a leaf may be used. The “length” of a leaf as used herein refers to the leaf dimension that is parallel to the leaf moving direction. The “width” of a leaf refers to the dimension of the leaf that is traverse the leaf moving direction and the direction of the radiation beam. The “height” of a leaf refers to the dimension of the leaf along the beam direction. The “side” of a leaf refers to the surface adjacent to neighboring leaves in a bank. The “end” of a leaf refers to the surface of the leaf inserted into the field along the length. FIG. 4 is a cross-sectional view of a portion of a multi level MLC 410 showing some detail of the leaf arrangement in accordance with some embodiments. As shown, first and second sets of leaves 420, 430 may be arranged at two different levels. In some embodiments, the first and second sets 420, 430 can be disposed such that each leaf in a set (e.g. leaf 422 in the first set 420) can be offset from a leaf in another set (e.g. leaf 432 in the second set 430) along a lateral direction or a direction traverse to the leaf moving direction. For example in some embodiments, a leaf in the first set 420 or second set 430 can be offset from a leaf in the second set 430 or first set 420 by substantially half a leaf. Alternatively, in some embodiments the first and second sets 420, 430 are disposed such that the gap between two adjacent leaves at a level (e.g. the gap between leaves 422, 424 in the first set 420) is positioned substantially at the middle of a leaf in another set (e.g. leaf 432 in the second set 430). The offset arrangement of leaves at different levels provides for leaf projections that are also offset at the isocenter. Therefore, in some embodiments the leaves in the first and second sets 420, 430 are arranged offset each other to provide for projections offset by approximately half a leaf width as projected at the isocenter plane. This provides for substantially an equivalent of doubling MLC definition, or improving the step resolution to half as compared to the definition of a single level MLC with leaves of the same physical width. In some embodiments, the MLC may include three or more sets of leaves at three or more levels which may be arranged such that each leaf at a level is offset e.g. by ⅓ or 1/n of leaf width as projected at the isocenter where n is the number of sets of the MLC. In embodiments with offset leaf arrangement, the number of leaves at a level may be different from the number of leaves at another level. For example, in a two level MLC, a leaf array at the upper level may include one more leaf than a leaf array at the lower level to ensure coverage by at least one single leaf at the sides of a symmetric MLC field. The leaves in a set at a level may have a substantially same cross-section. For example, in some embodiments the leaves in a set may have a same trapezoidal cross-section. Other cross-sectional shapes of leaves such as rectangular shape, tilted trapezoids, or trapezoids with stepped or wavy sides are possible. Alternating patterns of cross-sections are also possible, such as trapezoid, rectangle, trapezoid, rectangle, and so on. The cross-sectional shapes described herein do not refer to additional detail features in the cross-section that provide support and guidance for the leaves, such as added hook or tab shapes. Due to divergence of radiation beam from a source, the physical width of leaves at different levels may be different to provide the same projected width definition at the isocenter. For example, the leaves in a set closer to a source may have a narrower cross section than that of the leaves in a set farther from the source. In some embodiments, the leaves in the first and second sets may be arranged tilted or inclined to the source or substantially focus on a converging virtual point located substantially at the source. The focused leaf arrangement may improve the quality of beam shaping at the isocenter. The leaf side surfaces may be flat. In some embodiments, the adjacent leaf side surfaces may form a gap or spacing ranging from approximately 10 to 100 micro-meters to facilitate relative movement between the leaves. The leaf side gaps may be substantially the same at a level. Because leaves at a level may cover radiation leakage between leaf sides (gap leakage) at another level, the leaf sides of the multi level MLC of the invention require little or no “tongue in groove” design as in conventional MLCs. In some embodiments, the leaves may have a trapezoidal cross section and the leaves may be arranged such that the leaf side surfaces substantially focus on a converging virtual point located substantially at the radiation source. This arrangement may provide the least leaf side penumbra. This arrangement can also eliminate or minimize “tongue and groove effect” because at a leaf level there is substantially no leaf material overlap between leaves as viewed from the radiation source. In a real situation where the radiation source is of finite size, rather than a theoretical point, the radiation may be thought to emanate from various “pixels” within that finite source and, the leaf side surfaces may not be viewed as perfectly focused from every source pixel, and leaf overlap material at a leaf level may contribute a slight tongue and groove effect from some of those pixels. Rather than an ideal focus on the source, a practical compromise such as a small step, wave, or a very slight defocused tilt may prove to be a better balance between gap leakage and tongue and groove effects. The leaf ends can be round, flat, or in various other configurations. The penumbra of leaf ends closer to the source will tend to be greater than the penumbra of leaf ends farther from the source due to geometric projection effects of the radiation source. For treatment planning purposes, it would be desirable if leaf ends have approximately same penumbra. For leaf ends that are substantially rounded, the otherwise worse penumbra of the upper leaves can be partially mitigated by using a larger leaf end radius. A larger radius reduces the penumbra due to transmission through the leaf material (e.g. tungsten). A larger leaf end radius may require taller leaves. Therefore, in some embodiment the height of the upper level leaves is greater than the height of the lower level leaves to insure approximately constant penumbra over the entire leaf travel range. In some embodiments the height of the upper level leaves and the lower level leaves is substantially the same, but the upper level leaves have an end portion with a larger radius. FIG. 5 shows with greater detail the end portion of the upper level leaves in an exemplary MLC 510. The MLC 510 includes upper level leaves 512 and lower level leaves 514 supported by a support body such as a leaf box 516. The upper leaves 512 and lower level leaves 514 may have a main portion of substantially same height (“H” as shown). The end portion 518 of the upper level leaves 512 may have one or two “tooth” portions or projections 520a, 520b extended e.g. either upward or downward, or extended both upward and downward to allow an increase of the leaf end radius of the upper leaves 512. The extended radius can mitigate the penumbra of the upper leaves 512 without substantially increasing the weight or height of the upper level leaf bodies. Not increasing the leaf body weight beyond what is needed for shielding is desirable due to packaging volume and leaf weight constraints. The tooth extensions 520a, 520b may be located outside of the leaf box 516 and use the space not otherwise needed. The resulting upper leaves 512 may have an end portion 518 with a “mushroom” shape in side view. If necessary to further mitigate unequal leaf end penumbra between the upper and lower levels, the radius of the lower leaf ends can be reduced below the maximum radius allowed by the leaf height. The leaves may be constructed with various suitable radiation attenuating materials. To generally improve on leakage performance of existing beam limiting devices, the combined attenuation of all levels of the MLC should be approximately 2.5 tenth value layers (“TVLs”) or greater. Single leaves at one level should substantially mitigate the local leakage of leaf gaps at another level. In general, the leaf gap leakage that can be allowed at a level in the multi level MLC can be greater as compared to conventional single level MLCs since the leaves at another level can mitigate the gap leakage. Since small areas at the prescribed boundaries of treatment fields may be covered by only a single leaf, the leaf height should be 1.5 TVLs or greater to perform adequately. In some embodiments, the multi level MLC of the invention can provide for a treatment field that is shaped by leaves all having the same width definition at the isocenter. By way of example, a treatment field of 40×40 cm2 with a projected leaf width of ½ cm (¼ cm offset definition) can be provided using 322 individual leaves disposed at two levels. As another example, a treatment field of 40×40 cm2 with a projected leaf width of 1 cm (½ cm offset definition) can be provided using 162 individual leaves. It will be appreciated that treatment fields of different sizes with different width definitions can be provided by the multi level MLCs of the invention including different numbers of individual leaves based on specific applications. In some embodiments, the multi level MLC of the invention may provide for a treatment field that is shaped by leaves of different width definitions at the isocenter. The finer definition (e.g. ¼ cm) may be provided in the central portion of the treatment field where precision is more needed. This may reduce MLC cost and increase MLC reliability compared to an MLC with a greater number of leaves allowing fine definition throughout the entire treatment field. In an embodiment, the transition of leaf width can be gradual. For example, the width of leaves at a level can be progressively increased with distance from the center of the treatment field. Each leaf at a level may have a physically different width dimension. Alternatively, each MLC level may include leaf sections so that the transition of leaf widths is discreet. The transition can be made by placing transition leaves at specific locations on one or both levels. The transition leaves insure that the gaps between leaves project at the desired spacing for the desired definition regions. FIG. 6A is a cross-sectional view of a portion of an exemplary multi level MLC 610 providing variable width definition in accordance with some embodiments. The MLC 610 may include two or more sets of leaves 620, 630 of different sizes which project different leaf widths at the isocenter (e.g., ½, 1, 2 cm etc.). To simplify description, leaves at a level are shown as having a rectangular cross-section to better illustrate the offset arrangement of leaves at different levels. Leaves may have a cross-section of trapezoidal, rectangular, or other shapes. At a first level 620, the MLC 610 may include a first section of leaves 622 with a first cross-section that provides for a first substantially same width definition (e.g. ½ cm), a second section of leaves 624 with a second cross-section that provides for a second substantially same width definition (e.g. 1 cm), and optionally a third section of leaves 626 with a third cross-section that provides for a third substantially same width definition (e.g. 2 cm) at the isocenter, and so on. At a second level 630, the MLC 610 may include first, second, and optionally third sections of leaves 632, 634, 636 which may be arranged offset from the corresponding first, second, and optionally third sections of leaves 622, 624, 626 at the first level 620. The leaves of the first, second, and optionally third sections 632, 634, 636 at the second level 630 may have cross-sections that provide for width definitions at the isocenter substantially same as the first, second, and optionally third width definitions of the first level leaves 620 respectively. At one or both level(s) such as the first level 620, one or more transition leaves 627 may be disposed between the first and second sections 622, 624, or optionally one or more transition leaves 628 between the second and third sections 624, 626. By way of example, the first sections of leaves 622, 632 at the first and second levels 620, 630 may provide for ½ cm width definition or ¼ cm offset definition at the isocenter, the second sections of leaves 624, 634 at the first and second levels 620, 630 may provide for 1 cm width definition or ½ cm offset definition, and optionally the third sections of leaves 626, 636 at the first and second levels 620, 630 may provide for 2 cm width definition or 1 cm offset definition. In some embodiments, a transition leaf 627 may provide for ¾ cm width definition, or optionally a transition leaf 628 may provide for 1½ cm width definition. It should be noted that the above leaf width dimensions are provided by way of example, and it will be appreciated that different width definitions may be provided for by a multi level MLC 610 including leaves of different sizes. A multi level MLC with variable width definitions allows the use of different types or sizes of leaves in the MLC. For example, the MLC may include high definition leaves in the middle section to define a treatment field closely conformal to the target. In the outer section where high definition may not be required, relatively low definition leaves may be used to reduce manufacturing cost and increase reliability of the MLC. By way of example, a multi level MLC with a variable leaf width configuration illustrated in FIG. 6A may provide a 40×40 cm2 treatment field using only 162 leaves, which are far fewer than 322 leaves that would be required for ¼ cm definition across the full field. FIG. 6B illustrates another alternative embodiment with variable leaf widths which can provide a 40×40 cm2 treatment field also with 10 cm of ¼ cm definition using 202 leaves. In some aspect the invention provides for a method of shaping radiation beams. Using a multi level MLC and a control method provided by the invention, various radiation leakage can be significantly reduced. The leakage between leaf sides or gap leakage can be mitigated by using a multi level MLC with offset leaf arrangement (see FIGS. 3 and 6A-6B). As described above, a multi level MLC may include two or more sets of leaves at different levels, and leaves at each level may be arranged in two banks or arrays forming a plurality of pairs of opposing leaves at each level. The two or more sets of leaves may be disposed generally in parallel so that all the leaves of the multi level MLC may travel in a substantially same direction generally traverse to the beam direction. In a preferred embodiment, the two or more sets of leaves can be disposed such that leaves at a level are offset from leaves at another level in a lateral direction (e.g., y-direction) generally traverse to the leaf moving direction so that the leakage between leaf sides at a level can be mitigated by leaves at another level. To reduce leakage between abutted leaf ends that may be intended to close in shaping a treatment field, the ends of the abutted leaves at a level may close at a position slightly offset, in the leaf travel direction (e.g., x-direction), from the position where the ends of the abutted leaves close at a different level. This would ensure that the abutted leaf end leakages are not superimposed but instead attenuated by at least a single leaf height. FIG. 7A schematically shows an exemplary embodiment where abutted leaf ends close at positions 702, 704, which are offset in the leaf travel direction (e.g. the x-direction) so that the rays through pairs of abutted ends are never superimposed. In execution, factors such as 3-dimensional effects including the presence of separated treatment field regions the relative x and y positions of their field boundaries, and whether an even or odd number of field strips separate regions etc. should be accounted for in determining the offset positions. In general as shown in FIG. 7B, the abutted leaf end leakage can be mitigated if a portion of a leaf 712 in a leaf bank at a level overlaps a portion of a leaf 714 in an opposing leaf bank at a different level as viewed from a radiation source. This would allow mitigation of abutted leaf end leakage to acceptable levels without ever having to touch the opposing abutted leaves together. For example, a minimum physical gap of less than 1 mm between abutted leaf ends should sufficiently control leakage, yet still be manageable within control accuracies. Not requiring abutted leaf ends to ever actually touch can reduce control program and leaf drive mechanical complexity and increase leaf drive reliability. Components such as springs and sacrificial “fuses” in a leaf drive nut as used in conventional MLCs to limit collision damage can also be eliminated if abutted leaf contact is not needed and such collisions become a rare event. Penumbra compromises associated with complex interlocking leaf end shapes can also be avoided. The ability to dynamically close leaf ends quickly, with low leakage, even between momentarily separated field regions can be an advantage to treatment planning. Such a creative offset control can be applied to dynamically changing field regions. Dynamically separating and recombining field regions can be created generally without even momentarily producing an unwanted region of high abutted leaf end leakage. A multi level MLC and a method of shaping radiation beams have been described. One of the advantages of the multi level MLC is that the offset arrangement of leaves can effectively improve beam shaping resolution, and allow the same definition with leaves physically twice as wide as for a single level MLC. The extra physical leaf width is a considerable construction advantage for achieving equal or higher MLC definition in a more limited volume, particularly for screw leaf drive systems. For example, in screw leaf drive systems, long slender leaf drive screws may be susceptible to column buckling in a way that scales dramatically worse with smaller screw diameters. Since the leaf drive screw diameter is generally limited to not be greater than physical leaf width, the invention greatly reduces the screw drive susceptibility to column buckling by allowing leaf drive screws to be nearly doubled in diameter. In addition, wider leaves allow room for larger diameter motors. The general relaxation of leaf drive miniaturization can also allow more motor choices, faster leaf speeds, better manufacturing process control, higher performance margins, higher reliability, and easier service access. These advantages are all desirable for dynamic treatments and MLC cost is also reduced. Another advantage of the invention is that the use of the multi level MLC can significantly improve the leakage effect over a single level MLC used in conjunction with one or two pairs of collimation jaws. FIGS. 8A-8C and FIGS. 9A-9C compare a conventional beam shaping method with some embodiments of the invention and their leakage effect. The gray tones of the figure approximate the transmission of the radiation beam passing through the MLC similar to how it would appear on film, with more radiation intensity being darker. To simplify calculations in this example, each level provides 2 Tenth Value Layer (TVL) attenuation. Thus the transmission of radiation through a single leaf for this example is assumed to be 1% of the intensity of the original unattenuated radiation beam. FIG. 8C shows intended radiation field regions shaped by corresponding positions of a pair of collimation jaws (FIG. 8A) and a single level MLC (FIG. 8B) acting in combination. FIG. 9C shows intended radiation field regions shaped by a multi level MLC of the invention including a first set (FIG. 9A) and a second set (FIG. 9B). In the conventional method, the combined leakage between the abutted leaf ends separating field regions and the leakage between leaf sides are evident as shown in FIG. 8C, whereas in the method using a multi level MLC of the invention, the combined leakage is significantly reduced as shown in FIG. 9C. Further, the combined leakage reduction of a conventional MLC shown in FIG. 8C is limited to a rectangle, while the combined leakage reduction of the multi level MLC of the invention extends nearly to the boundaries of the treatment field shown in FIG. 9C. The leakage between abutted leaf ends can also be significantly mitigated using the control method described above. FIGS. 10A-10C show that the abutted leaf leakage between rounded leaf ends at a single level can be as much as 24% on the centerline. With the offset control in the leaf travel-direction between levels, the abutted leaf end leakage can be reduced to less than 1%, as shown in FIG. 10C. Treatment field regions can be quickly separated and recombined without high leakage. Because the multi level MLC and control method provided by the invention can effectively reduce leaf leakage to acceptable levels, collimation jaws such as y-direction jaws are not required to control leaf end to end leakage as it is in most conventional single level MLCs. A fairly small and lightweight y-direction jaw pair may optionally be used in conjunction with the multi level MLC to provide for continuous adjustability of field width. A y-direction jaw pair might also mitigate small points of leakage where the abutted leaf gap of one level aligns with a leaf side gap of the other level. Shielding fixed to carriages or to a unicarriage may be used to provide adequate TVL coverage under all use cases. FIG. 11 shows a side cross-sectional view of an exemplary two level MLC 1110 including upper level leaves 1112 and lower level leaves 1114 supported by a carriage 1116. A small shielding block 1118 can be fixed to the top of the carriage 1116 to insure adequate shielding under all use cases in combination with lowered leaf tail portions 1120, 1122. The multi level MLC of the invention can be used in a radiotherapy machine to support various treatment options including intensity-modulated radiation therapy (IMRT), arc therapy, and other forms of radiotherapy. In intensity-modulated radiation therapy, the multi level MLC can be controlled to modulate the intensity and adjust the shape of the beam conformal to the size, shape, and location of the target. In dynamic arc therapy, the radiation source may rotate in delivery of radiation from various angles. The multi level MLC can be dynamically controlled during rotation of the source to adjust the beam conformal to the size, shape, and location of the target from various angles. Those skilled in the art will appreciate that various other modifications may be made within the spirit and scope of the invention. All these or other variations and modifications are contemplated by the inventors and within the scope of the invention. |
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048572632 | summary | BACKGROUND OF THE INVENTION This invention relates to the nuclear-reactor art. In this art, the nuclear-reactor plants are provided with spent-fuel storage facilitates, specifically a spent-fuel storage pool having racks for storing spent fuel under a substantial depth of water. This invention relates particularly to the storage of spent fuel and to racks for storing such fuel. Nuclear plants produce large quantities of spent fuel. In the past, it was contemplated that the spent fuel would be reprocessed to provide fissionable uranium and plutonium as a renewed fuel. But the Nuclear Non-Proliferation Treaty, to which the United States is a party, has been interpreted to bar the reprocessing of fuel inasmuch as plutonium, a product of the reprocessing, is a weapons material. It then becomes necessary to cope with the problem of safely handling the spent fuel which continuously emerges from nuclear plants. It is an object of this invention to provide an effective solution for this problem. It is contemplated that the spent fuel must be stored for a number of years before it can safely be disposed of as nuclear waste. The solution of the spent-fuel problem deals with the storage of the spent fuel during these years. Spent fuel retains a measure of reactivity, i.e., neutron emissivity, which is appreciable but is insufficient for economic use in a reactor. It is then necessary that the spent fuel be stored in such a way that the mass stored does not become critical. In refueling a reactor, the fuel assemblies in specified areas of the reactor are replaced at intervals of several years. The residual reactivity of the removed fuel assemblies throughout each area of the refueling is not uniform. It is then necessary, in the storage of spent fuel, to preclude nuclear criticality by reason of the presence of fuel assemblies having high residual reactivity. In accordance with the teachings of the prior art, such criticality is precluded by providing racks whose cells are appropriately spaced. In addition, quantities of neutron-absorbing material or poison can be provided in the cells of the racks in which the spent fuel is stored. The first of these expedients requires that the volume occupied by racks be unreasonably large. The second expedient introduces a high cost factor. It is an object of this invention to provide for the storing of spent fuel in a way that shall be economic both financially and with respect to volume. This invention relates to storage of spent fuel in a rack of uniformly spaced cells. The cell dimensions and the actual center-to-center (CTC) spacing is set to accommodate the types of fuel which are to be stored in the rack. Each fuel-storage pool is subdivided into two regions, herein designated Region 1 and Region 2. Region 1 is the smaller region and is reserved for off-core loading; i.e., for temporary loading of spent-fuel assemblies, regardless of burn-up, as they are removed from the reactor. Typically, Region 1 may serve to load about 200 fuel assemblies at a low fuel-assembly density using a fraction, usually half of the available storage locations. Region 2 is reserved for storing, for the required long time interval, the assemblies from Region 1 which have been found to have sustained at least a minimum predetermined burn-up. As used in this application, the expression "storage location" means generally a location where fuel assemblies are stored. Specifically, this term comprehends cells or locations which may be formed between a plurality of cells. In Region 1 the storage locations are in a checkerboard pattern in a honeycomb type structure. One set of storage locations in this honeycomb are capped to prevent the insertion of fuel assemblies, while the other set is uncapped. In a typical checkerboard pattern, a square of one color, for example, black alternates with a square of another color, for example, red, so that the total area of black squares is equal to the total area of red squares. Typically, the checkerboard pattern or honeycomb structure of Region 1 in the practice of this invention may take this form. However, a structure in which a number of contiguous cells are capped in uniformly spaced areas of the cell surface may also be provided. This structure is necessary to reduce the possibility of criticality. The reference in this application to a "generally checkerboard" pattern is intended to cover this structure as well as one simulating an actual checkerboard. It is necessary that the reactivity K.sub.eff of the spent fuel assemblies in Region 1 be maintained at less than or equal to 0.95. For this purpose, the cells may be provided with neutron-absorbing material. The water in which the rack is immersed may be maintained adequately neutron absorbent by the solution therein of a boron compound or the compound of another neutron-absorbing element. The spent-fuel assemblies in Region 1 are surveyed through administrative control to determine burn-up. Those assemblies that have sustained the required burn-up are transferred to Region 2 where they remain until their radioactivity is reduced to a magnitude permitting removal from the pool and other disposal. It is desirable that, in the use of a spent-fuel storage pool, the option be afforded to include neutron-absorbing poison in the rack either before the rack is installed or after it is installed. In spent-fuel facilities in accordance with the teachings of the prior art, such racks are not provided. If neutron poison is necessary, it must be included before the rack is installed in the pool; it cannot readily be included in the rack after the rack is installed. Another deficiency of prior art spent-fuel facilities is that the CTC spacing between storage cells is maintained by grids or like components. This is an undesirable and space-consuming complication. It is an object of this invention to overcome the above-described disadvantages and drawbacks of the prior art and to provide a spent-fuel storage rack which shall occupy a minimum volume and shall afford the option of including the nuclear poison either before or, readily, after the rack is installed in the pool and in whose use the CTC spacing shall be reliably maintained with economy in space and without spacing grids. SUMMARY OF THE INVENTION In accordance with this invention a spent-fuel storage rack is provided which which occupies a minimum volume. The cells of this rack are each composed of a body formed of sheet metal, typically stainless steel sheet. The sheet for each cell is formed into the desired shape, for example, an elongated hollow body of generally polygonal transverse cross section or of circular or other cross section. The bodies are mounted on a base plate with each body except those along the periphery of the rack, secured, to bodies adjacent to it, along a welded joint common to the secured bodies. The joint is formed by welds spaced longitudinally at the joint along the secured bodies. The spacing for each weld is such as to optimize the natural frequency of the rack to minimize the response produced by seismic accelerations characteristic of the geographical area where the rack is to be installed. Where the bodies are of transverse polygonal cross section, each body is secured to the adjacent bodies along the apices formed by the sides of the bodies. It is desirable that the option be available to include neutron-absorbing poison either before or after the rack is installed. For this purpose, pockets for neutron-absorbing material are, in accordance with this invention, provided along the outer surface of each body which forms a cell. Each pocket is formed between the surface and a wrapper plate secured to the surface. Where the bodies are of polygonal section, a pocket is formed between each side of a body and the wrapper plate. An additional spent-fuel storage location is also formed between each body or cell and the sequential bodies or cells contiguous to it. This storage location has in common with the bodies between which it is formed the wall sections and pockets of the latter bodies. It is to be borne in mind that there is symmetry about each body within the rack and that the above selection of a body and the bodies secured to it is applicable to any set of bodies within the rack. Any body within the rack is at the same time a body to which adjacent bodies are secured, one of the secured bodies, or the additional bodies having walls common to the body having adjacent bodies secured to it and the sequential adjacent bodies. The pockets have additional advantages. They maintain the required CTC spacing between the assemblies which are placed in the spent-fuel storage locations without the necessity for grids. In addition they maintain fuel-assembly-to-fuel-assembly separation which permits closer spacing of the fuel assemblies where this spacing is limited by the necessity of avoiding criticality. The assemblies are placed in the cells so as to have limited movement; typically the spacing of an assembly from the walls of a cell is 1/4 inch. |
description | This application claims the benefit of priority to U.S. Provisional Application 62/521,692, filed Jun. 19, 2017, which is hereby incorporated herein by reference in its entirety. This invention was made with Government support under Grant No. DE-NE0008407 awarded by the Department of Energy. The Government has certain rights in the invention. There are many areas where a robust, long-term, low-maintenance power supply is needed, such as in deep sea exploration, interplanetary and interstellar exploration, cardiac pacemakers, polar explorations, and military equipment. The devices and methods discussed herein address these and other needs. In accordance with the purposes of the disclosed devices and methods, as embodied and broadly described herein, the disclosed subject matter relates to charge generating devices. The charge generating devices can comprise a substrate having a top surface and a bottom surface; a plurality of spaced-apart three-dimensional elements disposed on the top surface of the substrate; and a plurality of cavities formed by the plurality of spaced-apart three-dimensional elements, the plurality of cavities being the area between the plurality of spaced-apart three-dimensional elements. In some examples, the substrate and the plurality of spaced-apart three-dimensional elements can comprise a wide band-gap semiconductor, such as SiC, GaN, Ga2O3, diamond, GaAs, AlN, AlGaN, Al2O3, BN, AlAs, GaP, InP, B4C, or combinations thereof. In some examples, the substrate and the plurality of spaced-apart three-dimensional elements can comprise SiC, diamond, GaN, Ga2O3, or Al2O3. In some examples, the plurality of spaced-apart three-dimensional elements are integrally formed with the substrate. Each of the plurality of spaced-apart three-dimensional elements can have an average characteristic of from 0.5 μm to 500 μm. Each of the plurality of spaced-apart three-dimensional elements can have an average height of from 1 μm to 500 μm. In some examples, the plurality of spaced-apart three-dimensional elements can form an array and each of the plurality of spaced-apart three-dimensional elements is separated from its neighboring three-dimensional elements 108 by a distance of from 1 μm to 500 μm (edge to edge). For example, the plurality of spaced-apart three-dimensional elements can form a rectangular array, a hexagonal array, or a linear array. The devices can, in some examples, further comprise a radioactive layer disposed on at least a portion of the plurality of spaced-apart three-dimensional elements and the top surface such that the plurality of cavities and the top surface are substantially coated by the radioactive layer, thereby forming a plurality of coated cavities. The radioactive layer can, for example, emit alpha radiation, beta radiation, gamma radiation, or a combination thereof. In some examples, the radioactive layer emits alpha radiation. The radioactive layer can comprise a radioisotope. For example, the radioactive layer can comprise 148Gd, 238Pu, 208Po, 210Po, 243Cm, 244Cm, 241Am, 63Ni, 106Ru, 235U, 204Tl, 14C, 3H, 85Kr, 90Sr, 90Y, 147Pm, 109Gd, or a combination thereof. In some examples, the radioactive layer comprises 241Am. The radioactive layer can, for example, have a thickness of from 0.1 μm to 100 μm. The devices can further comprise a first conducting layer disposed above the plurality of spaced-apart three-dimensional elements, wherein the first conducting layer is in electric contact with at least a portion of the plurality of spaced-apart three-dimensional elements. The devices can further comprise a second conducting layer disposed below the substrate, wherein the second conducting layer is in electric contact with the bottom surface of the substrate. In some examples, the first conducting layer, the second conducting layer, or a combination thereof comprise(s) a metal. The metal can comprise, for example, a metal selected from the group consisting of Al, Ti, Ni, Cu, Ga, Ag, Ir, Pt, Au, Cr, Mo, Pd, W, and combinations thereof. In some examples, the first conducting layer, the second conducting layer, or a combination thereof can comprise(s) a radioisotope. For example, the first conducting layer, the second conducting layer, or a combination thereof comprise(s) 63Ni. The devices can further comprise a first scintillation layer disposed above the first conducting layer and in radiative contact with the first conducting layer. In some examples, the devices can further comprise a second scintillation layer disposed below the second conducting layer and in radiative contact with the second conducting layer. In some examples, the first scintillation layer can be in physical contact with the first conducting layer. In some examples, the second scintillation layer can be in physical contact with the second conducting layer. The first scintillation layer, the second scintillation layer, or a combination thereof can comprise(s) a scintillating material selected from the group consisting of a ZnS-based phosphor, NaI, CsI, cerium-doped lutetium yttrium orthosilicate (Ce:Lu1.8Y0.2SiO5, LYSO), bismuth germanate (Bi4Ge3O12, BGO), plastic, CeF3, europium doped calcium fluoride (Eu:CaF2), PbWO4, CdWO4, cerium doped cesium lithium yttrium chloride (Ce:Cs2LiYCl6, CLYC), Ce:LaCl3, CeBr3, Ce:LaBr3, praseodymium doped lutetium aluminum garnet (Pr:Lu3Al5O12, Pr:LuAG), cerium doped lutetium aluminum perovskite (Ce:LuAlO3, Ce:LuAP), Ce:Lu3Al5O12, cerium doped yttrium orthosilicate (Ce:Y2SiO5, Ce:YSO), cerium doped yttrium aluminum perovskite (Ce:YAlO3; Ce:YAP), cerium doped yttrium aluminum garnet (Ce:Y3Al5O12, Ce:YAG), and combinations thereof. In some examples, the device can further comprise a scintillating material disposed within at least a portion of the plurality of coated cavities; a layer of a scintillating material disposed on the radioactive layer, such that the plurality of coated cavities are substantially coated by the layer of scintillating material; or a combination thereof. The scintillating material can, for example, be selected from the group consisting of a ZnS-based phosphor, CsI, thallium-doped sodium iodide (Tl:NaI), cerium-doped lutetium yttrium orthosilicate (Ce:Lu1.8Y0.2SiO5, LYSO), bismuth germanate (Bi4Ge3O12, BGO), plastic, CeF3, europium doped calcium fluoride (Eu:CaF2), PbWO4, CdWO4, cerium doped cesium lithium yttrium chloride (Ce:Cs2LiYCl6, CLYC), Ce:LaCl3, CeBr3, Ce:LaBr3, praseodymium doped lutetium aluminum garnet (Pr:Lu3Al5O12, Pr:LuAG), cerium doped lutetium aluminum perovskite (Ce:LuAlO3, Ce:LuAP), Ce:Lu3Al5O12, cerium doped yttrium orthosilicate (Ce:Y2SiO5, Ce:YSO), cerium doped yttrium aluminum perovskite (Ce:YAlO3; Ce:YAP), cerium doped yttrium aluminum garnet (Ce:Y3Al5O12, Ce:YAG), and combinations thereof. In some examples, the scintillating material 124 can perform as an energy degrader. Also disclosed herein are charge generating devices comprising: a substrate having a top surface and a bottom surface; a plurality of spaced-apart three-dimensional elements disposed on the top surface of the substrate; a plurality of cavities formed by the plurality of spaced-apart three-dimensional elements, the plurality of cavities being the area between the plurality of spaced-apart three-dimensional elements; a radioactive material disposed within at least a portion of the plurality of cavities, thereby forming a plurality of at least partially filled cavities; a first conducting layer disposed above the plurality of at least partially filled cavities and the plurality of spaced-apart three-dimensional elements, wherein the first conducting layer is in electric contact with the plurality of spaced-apart three-dimensional elements; a second conducting layer disposed below the substrate, wherein the second conducting layer is in electric contact with the bottom surface of the substrate; a first scintillation layer disposed above the first conducting layer and in radiative contact with the first conducting layer; and a second scintillation layer disposed below the second conducting layer and in radiative contact with the second conducting layer. In some examples, the plurality of spaced-apart three-dimensional elements are integrally formed with the substrate. Also disclosed herein are charge generating devices comprising: a substrate having a top surface and a bottom surface; a plurality of spaced-apart three-dimensional elements disposed on the top surface of the substrate; a plurality of cavities formed by the plurality of spaced-apart three-dimensional elements, the plurality of cavities being the area between the plurality of spaced-apart three-dimensional elements; a scintillating material disposed within at least a portion of the plurality of cavities, thereby forming a plurality of at least partially filled cavities; a first conducting layer disposed above the plurality of at least partially filled cavities and the plurality of spaced-apart three-dimensional elements, wherein the first conducting layer is in electric contact with the plurality of spaced-apart three-dimensional elements; a second conducting layer disposed below the substrate, wherein the second conducting layer is in electric contact with the bottom surface of the substrate; a first scintillation layer disposed above the first conducting layer and in radiative contact with the first conducting layer; and a second scintillation layer disposed below the second conducting layer and in radiative contact with the second conducting layer. In some examples, the plurality of spaced-apart three-dimensional elements are integrally formed with the substrate. In some examples, the scintillating material is disposed as a layer on the plurality of spaced-apart three-dimensional elements, such that the first conducting layer is in electric contact with the plurality of spaced-apart three-dimensional elements through the layer of the scintillating material. In some examples, the device can further comprise a radioactive layer disposed on at least a portion of the plurality of spaced-apart three-dimensional elements and the top surface such that the plurality of cavities and the top surface are substantially coated by the radioactive layer, thereby forming a plurality of coated cavities such that the scintillating material is disposed within at least a portion of the plurality of coated cavities. In some examples, the devices can further comprise a charge storage device electrically coupled to the first conducting layer and the second conducting layer. Also disclosed herein are methods of making the devices disclosed herein. For example, the methods can comprise forming the plurality of spaced-apart three-dimensional elements on the top surface of the substrate; and depositing the radioactive layer on at least a portion of the plurality of spaced-apart three-dimensional elements and the top surface such that the plurality of cavities and the top surface are substantially coated by the radioactive layer, thereby forming the plurality of coated cavities. In some examples, wherein the device further comprises a scintillating material disposed within at least a portion of the plurality of coated cavities, the method further comprises depositing the scintillating material within at least a portion of the plurality of coated cavities. In some examples, therein the scintillating material comprises a radioactive scintillating material, the method further comprises depositing the radioactive scintillating material within at least a portion of the plurality of coated cavities. In some examples, the methods of making the device can comprise forming the plurality of spaced-apart three-dimensional elements on the top surface of the substrate; and depositing the scintillating material within at least a portion of the plurality of cavities. Additional advantages of the disclosed devices and methods will be set forth in part in the description which follows, and in part will be obvious from the description. The advantages of the disclosed devices and methods will be realized and attained by means of the elements and combinations particularly pointed out in the appended claims. It is to be understood that both the foregoing general description and the following detailed description are exemplary and explanatory only and are not restrictive of the disclosed devices and methods, as claimed. The details of one or more embodiments of the invention are set forth in the accompanying drawings and the description below. Other features, objects, and advantages of the invention will be apparent from the description and drawings, and from the claims. Devices Discussed herein are charge generating devices (e.g., electricity generating devices). Referring now to FIG. 1, in some examples the devices 100 can comprise a substrate 102 having a top surface 104 and a bottom surface 106; a plurality of spaced-apart three-dimensional elements 108 disposed on the top surface 104 of the substrate 102; and a plurality of cavities 110 formed by the plurality of spaced-apart three-dimensional elements 108, the plurality of cavities 110 being the area between the plurality of spaced-apart three-dimensional elements 108. The substrate 102 and the plurality of spaced-apart three-dimensional elements 108 can be fabricated from any suitable material or combination of materials compatible with the devices and methods described herein. In some examples, the substrate 102 and the plurality of spaced-apart three-dimensional elements 108 comprise a wide band-gap semiconductor. Examples of suitable materials for the substrate 102 and the plurality of spaced-apart three-dimensional elements 108 include, but are not limited to, SiC, GaN, Ga2O3, diamond, GaAs, AlN, AlGaN, Al2O3, BN, AlAs, GaP, InP, B4C, and combinations thereof. In some examples, the substrate 102 and the plurality of spaced-apart three-dimensional elements 108 can comprise SiC, diamond, GaN, Ga2O3, or Al2O3. In some examples, the plurality of spaced-apart three-dimensional elements 108 are integrally formed with the substrate 102. The plurality of spaced-apart three-dimensional elements 108 can comprise three-dimensional elements with a cross-section in the plane of the first surface 104 of any regular or irregular shape (e.g., a circle, an ellipse, a triangle, a square, a rectangle, a hexagon, an octagon, a star, a polygon, an irregular shape, etc.). In some examples, the cross-section in the plane of the first surface 104 of plurality of spaced-apart three-dimensional elements 108 can be an isotropic shape. In some examples, the cross-section in the plane of the first surface 104 of plurality of spaced-apart three-dimensional elements 108 can be an anisotropic shape. Each of the plurality of spaced-apart three-dimensional elements 108 can have an average characteristic dimension. The term “characteristic dimension,” as used herein, refers to the largest straight line distance spanning a spaced-apart three-dimensional element 108 in the plane of the first surface 104. For example, in the case of a spaced-apart three-dimensional element 108 having a substantially circular shape in the plane of the first surface 104, the characteristic dimension of the spaced-apart three-dimensional element 108 is the diameter of the spaced-apart three-dimensional element. “Average characteristic dimension” and “mean characteristic dimension” are used interchangeably herein, and generally refer to the statistical mean characteristic dimension of the spaced-apart three-dimensional elements in a population of spaced-apart three-dimensional elements. The characteristic dimension can be measured using methods known in the art, such as evaluation by optical microscopy, scanning electron microscopy, transmission electron microscopy, and/or atomic force microscopy. In some examples, each of the plurality of spaced-apart three-dimensional elements 108 can have an average characteristic dimension of 0.5 micrometers (μm) or more (e.g., 1 μm or more, 2 μm or more, 3 μm or more, 4 μm or more, 5 μm or more, 10 μm or more, 15 μm or more, 20 μm or more, 25 μm or more, 30 μm or more, 40 μm or more, 50 μm or more, 60 μm or more, 70 μm or more, 80 μm or more, 90 μm or more, 100 μm or more, 125 μm or more, 150 μm or more, 175 μm or more, 200 μm or more, 225 μm or more, 250 μm or more, 300 μm or more, 350 μm or more, or 400 μm or more). In some examples, each of the plurality of spaced-apart three-dimensional elements 108 can have an average characteristic dimension of 500 μm or less (e.g., 450 μm or less, 400 μm or less, 350 μm or less, 300 μm or less, 250 μm or less, 225 μm or less, 200 μm or less, 175 μm or less, 150 μm or less, 125 μm or less, 100 μm or less, 90 μm or less, 80 μm or less, 70 μm or less, 60 μm or less, 50 μm or less, 40 μm or less, 30 μm or less, 25 μm or less, 20 μm or less, 15 μm or less, 10 μm or less, or 5 μm or less). The average characteristic dimension of each of the plurality of spaced-apart three-dimensional elements 108 can range from any of the minimum values described above to any of the maximum values described above. For example, each of the plurality of spaced-apart three-dimensional elements 108 can have an average characteristic dimension of from 0.5 μm to 500 μm (e.g., from 0.5 μm to 250 μm, from 250 μm to 500 μm, from 0.5 μm to 100 μm, from 100 μm to 200 μm, from 200 μm to 300 μm, from 300 μm to 400 μm, from 400 μm to 500 μm, or from 2 μm to 400 μm). Each of the plurality of spaced-apart three-dimensional elements 108 can have an average height. As used herein, the height of a three-dimensional element 108 refers to the largest straight line distance spanning the three-dimensional element 108 in a plane perpendicular to the first surface 104. For example, in the case of a three-dimensional element 108 having a substantially cylindrical shape, with the cross-section of the three-dimensional element 108 having a circular shape in the plane of the first surface 104, the height of the three-dimensional element 108 is the height of the cylindrical three-dimensional element 108. “Average height” of a plurality of spaced-apart three-dimensional elements 108 and “mean thickness” of a plurality of spaced-apart three-dimensional elements 108 are used interchangeably herein, and generally refer to the statistical mean thickness of the three-dimensional elements 108 in a population of three-dimensional element 108. The height of the plurality of spaced-apart three-dimensional elements 108 can be measured using methods known in the art, such as evaluation by optical microscopy, scanning electron microscopy, transmission electron microscopy, and/or atomic force microscopy. For example, each of the plurality of spaced-apart three-dimensional elements 108 can have an average height of 1 μm or more (e.g., 2 μm or more, 3 μm or more, 4 μm or more, 5 μm or more, 6 μm or more, 7 μm or more, 8 μm or more, 9 μm or more, 10 μm or more, 15 μm or more, 20 μm or more, 25 μm or more, 30 μm or more, 35 μm or more, 40 μm or more, 45 μm or more, 50 μm or more, 55 μm or more, 60 μm or more, 65 μm or more, 70 μm or more, 75 μm or more, 80 μm or more, 85 μm or more, 90 μm or more, 100 μm or more, 110 μm or more, 120 μm or more, 130 μm or more, 140 μm or more, 150 μm or more, 160 μm or more, 170 μm or more, 180 μm or more, 190 μm or more, 200 μm or more, 225 μm or more, 250 μm or more, 275 μm or more, 300 μm or more, 325 μm or more, 350 μm or more, 375 μm or more, 400 μm or more, 425 μm or more, or 450 μm or more). In some examples, each of the plurality of spaced-apart three-dimensional elements 108 can have an average height of 500 μm or less (e.g., 475 μm or less, 450 μm or less, 425 μm or less, 400 μm or less, 375 μm or less, 350 μm or less, 325 μm or less, 300 μm or less, 275 μm or less, 250 μm or less, 225 μm or less, 200 μm or less, 190 μm or less, 180 μm or less, 170 μm or less, 160 μm or less, 150 μm or less, 140 μm or less, 130 μm or less, 120 μm or less, 110 μm or less, 100 μm or less, 95 μm or less, 90 μm or less, 85 μm or less, 80 μm or less, 75 μm or less, 70 μm or less, 65 μm or less, 60 μm or less, 55 μm or less, 50 μm or less, 45 μm or less, 40 μm or less, 35 μm or less, 30 μm or less, 25 μm or less, 20 μm or less, 15 μm or less, 10 μm or less, 9 μm or less, 8 μm or less, 7 μm or less, 6 μm or less, or 5 μm or less). The average height of each of the plurality of spaced-apart three-dimensional elements 108 can range from any of the minimum values described above to any of the maximum values described above. For example, each of the plurality of spaced-apart three-dimensional elements 108 can have an average height of from 1 μm to 500 μm (e.g., from 1 μm to 250 μm, from 250 μm to 500 μm, from 1 μm to 100 μm, from 100 μm to 200 μm, from 200 μm to 300 μm, from 300 μm to 400 μm, from 400 μm to 500 μm, from 1 μm to 400 μm, from 1 μm to 300 μm, from 1 μm to 200 μm, from 1 μm to 50 μm, from 50 μm to 100 μm, from 1 μm to 20 μm, from 20 μm to 40 μm, from 40 μm to 60 μm, from 60 μm to 80 μm, from 80 μm to 100 μm, or from 5 μm to 90 μm). In some examples, the average characteristic dimension of each of the plurality of spaced-apart three-dimensional elements 108 can be substantially the same along the height of each of the plurality of spaced-apart three-dimensional elements 108. In some examples, the plurality of spaced-apart three-dimensional elements 108 can have a tapered profile. For example, each of the plurality of spaced-apart three-dimensional elements 108 can have an upper portion and a lower portion, wherein the lower portion of each of the plurality of spaced-apart three dimensional elements 108 is positioned closer to the first surface 104 of the substrate 102 relative to the upper portion, where the average characteristic dimension of the upper portion is different than the average characteristic dimension of the lower portion. For example, the average characteristic dimension of the upper portion can be less than the average characteristic dimension of the lower portion. An example of a device where the plurality of spaced-apart three-dimensional elements 108 have a tapered profile is shown in FIG. 2. Each of the plurality of spaced-apart three-dimensional elements 108 can have an aspect ratio. As used herein, the aspect ratio of each of the plurality of spaced-apart three-dimensional elements 108 is the ratio of the average characteristic dimension to the average height the three-dimensional element 108. For example, each of the plurality of spaced-apart three-dimensional elements 108 can have an aspect ratio of 1:200 or more (e.g., 1:175 or more, 1:150 or more, 1:125 or more, 1:100 or more, 1:75 or more, 1:50 or more, 1:25 or more, 1:10 or more, 1:5 or more, 1:4 or more, 1:3 or more, 1:2 or more, 1:1 or more, 2:1 or more, 3:1 or more, 4:1 or more, 5:1 or more, 10:1 or more, 25:1 or more, 50:1 or more, 75:1 or more, 100:1 or more, 125:1 or more, 150:1 or more, 175:1 or more, 200:1 or more, 250:1 or more, 300:1 or more, 350:1 or more, or 400:1 or more). In some examples, each of the plurality of spaced-apart three-dimensional elements 108 can have an aspect ratio of 500:1 or less (e.g., 450:1 or less, 400:1 or less, 350:1 or less, 300:1 or less, 250:1 or less, 200:1 or less, 175:1 or less, 150:1 or less, 125:1 or less, 100:1 or less, 75:1 or less, 50:1 or less, 25:1 or less, 10:1 or less, 5:1 or less, 4:1 or less, 3:1 or less, 2:1 or less, 1:1 or less, 1:2 or less, 1:3 or less, 1:4 or less, 1:5 or less, 1:10 or less, 1:25 or less, 1:50 or less, 1:75 or less, 1:100 or less, 1:125 or less, or 1:150 or less). The aspect ratio of each of the plurality of spaced-apart three-dimensional elements 108 can range from any of the minimum values described above to any of the maximum values described above. For example, each of the plurality of spaced-apart three-dimensional elements 108 can have an aspect ratio of from 1:200 to 500:1 (e.g., from 1:200 to 1:1, from 1:1 to 500:1, from 1:200 to 2:1. from 5:1 to 500:1, from 1:200 to 200:1, from 1:100 to 100:1, or from 1:150 to 400:1). In some examples, the plurality of spaced-apart three-dimensional elements 108 can form an array and each of the plurality of spaced-apart three-dimensional elements 108 is separated from its neighboring three-dimensional elements 108 by a distance of 1 μm or more (edge to edge) (e.g., 2 μm or more, 3 μm or more, 4 μm or more, 5 μm or more, 6 μm or more, 7 μm or more, 8 μm or more, 9 μm or more, 10 μm or more, 15 μm or more, 20 μm or more, 25 μm or more, 30 μm or more, 35 μm or more, 40 μm or more, 45 μm or more, 50 μm or more, 55 μm or more, 60 μm or more, 65 μm or more, 70 μm or more, 75 μm or more, 80 μm or more, 85 μm or more, 90 μm or more, 100 μm or more, 110 μm or more, 120 μm or more, 130 μm or more, 140 μm or more, 150 μm or more, 160 μm or more, 170 μm or more, 180 μm or more, 190 μm or more, 200 μm or more, 225 μm or more, 250 μm or more, 275 μm or more, 300 μm or more, 325 μm or more, 350 μm or more, 375 μm or more, 400 μm or more, 425 μm or more, or 450 μm or more). In some examples, the plurality of spaced-apart three-dimensional elements 108 can form an array and each of the plurality of spaced-apart three-dimensional elements 108 is separated from its neighboring three-dimensional elements 108 by a distance of 500 μm or less (e.g., 475 μm or less, 450 μm or less, 425 μm or less, 400 μm or less, 375 μm or less, 350 μm or less, 325 μm or less, 300 μm or less, 275 μm or less, 250 μm or less, 225 μm or less, 200 μm or less, 190 μm or less, 180 μm or less, 170 μm or less, 160 μm or less, 150 μm or less, 140 μm or less, 130 μm or less, 120 μm or less, 110 μm or less, 100 μm or less, 95 μm or less, 90 μm or less, 85 μm or less, 80 μm or less, 75 μm or less, 70 μm or less, 65 μm or less, 60 μm or less, 55 μm or less, 50 μm or less, 45 μm or less, 40 μm or less, 35 μm or less, 30 μm or less, 25 μm or less, 20 μm or less, 15 μm or less, 10 μm or less, 9 μm or less, 8 μm or less, 7 μm or less, 6 μm or less, or 5 μm or less). The distance each of the plurality of spaced-apart three-dimensional elements 108 is separated from its neighboring three-dimensional elements 108 in the array can range from any of the minimum values described above to any of the maximum values described above. For example, the plurality of spaced-apart three-dimensional elements 108 can form an array and each of the plurality of spaced-apart three-dimensional elements 108 is separated from its neighboring three-dimensional elements 108 by a distance of from 1 μm to 500 μm (edge to edge) (e.g., from 1 μm to 250 μm, from 250 μm to 500 μm, from 1 μm to 100 μm, from 100 μm to 200 μm, from 200 μm to 300 μm, from 300 μm to 400 μm, from 400 μm to 500 μm, from 1 μm to 400 μm, from 1 μm to 300 μm, from 1 μm to 200 μm, from 1 μm to 50 μm, from 50 μm to 100 μm, from 1 μm to 20 μm, from 20 μm to 40 μm, from 40 μm to 60 μm, from 60 μm to 80 μm, from 80 μm to 100 μm, or from 5 μm to 90 μm). For example, the plurality of spaced-apart three-dimensional elements 108 can form a rectangular array (e.g., as shown in FIG. 3), a hexagonal array (e.g., as shown in FIG. 4 and FIG. 5), or a linear array (e.g., as shown in FIG. 6 and FIG. 7). Top down views (e.g., looking down at the top surface 104 of the substrate 102) of devices 100 showing exemplary arrays formed by the plurality of spaced-apart three-dimensional elements 108 and the plurality of cavities 110 are shown in FIG. 3-FIG. 7, the gray elements represent the plurality of cavities 110 and the white elements represent the plurality of spaced-apart three-dimensional elements 108. The plurality of cavities 110 formed by the plurality of spaced-apart three-dimensional elements 108 can have a cross-section in the plane of the first surface 104 of any regular or irregular shape (e.g., a circle, an ellipse, a triangle, a square, a rectangle, a hexagon, an octagon, a star, a polygon, an irregular shape, etc.). In some examples, the cross-section in the plane of the first surface 104 of plurality of spaced-apart three-dimensional elements 108 can be an isotropic shape. In some examples, the cross-section in the plane of the first surface 104 of plurality of spaced-apart three-dimensional elements 108 can be an anisotropic shape. In some examples, the plurality of cavities 110 formed by the plurality of spaced-apart three-dimensional elements 108 can comprise a plurality of troughs. The devices can, in some examples, further comprise a radioactive layer 112 disposed on at least a portion of the plurality of spaced-apart three-dimensional elements 108 and the top surface 104 such that the plurality of cavities 110 and the top surface 104 are substantially coated by the radioactive layer 112, thereby forming a plurality of coated cavities 114. The radioactive layer 112 can, for example, emit alpha radiation, beta radiation, gamma radiation, or a combination thereof. In some examples, the radioactive layer 112 emits alpha radiation. The radioactive layer 112 can comprise a radioisotope. For example, the radioactive layer 112 can comprise 148Gd, 238Pu, 208Po, 210Po, 243Cm, 244Cm, 241Am, 63Ni, 106Ru, 235U, 204Tl, 14C, 3H, 85Kr, 90Sr, 90Y, 147Pm, 109Gd, or a combination thereof. In some examples, the radioactive layer 112 comprises 241Am. The radioactive layer 112 can, for example, have a thickness of 0.1 μm or more (e.g., 0.25 μm or more, 0.5 μm or more, 0.75 μm or more, 1 μm or more, 1.25 μm or more, 1.5 μm or more, 1.75 μm or more, 2 μm or more, 2.5 μm or more, 3 μm or more, 3.5 μm or more, 4 μm or more, 4.5 μm or more, 5 μm or more, 5.5 μm or more, 6 μm or more, 6.5 μm or more, 7 μm or more, 7.5 μm or more, 8 μm or more, 8.5 μm or more, 9 μm or more, 9.5 μm or more, 10 μm or more, 11 μm or more, 12 μm or more, 13 μm or more, 14 μm or more, 15 μm or more, 16 μm or more, 17 μm or more, 18 μm or more, 19 μm or more, 20 μm or more, 25 μm or more, 30 μm or more, 35 μm or more, 40 μm or more, 45 μm or more, 50 μm or more, 55 μm or more, 60 μm or more, 65 μm or more, 70 μm or more, 75 μm or more, 80 μm or more, 85 μm or more, or 90 μm or more). In some examples, the radioactive layer 112 can have a thickness of 100 μm or less (e.g., 95 μm or less, 90 μm or less, 85 μm or less, 80 μm or less, 75 μm or less, 70 μm or less, 65 μm or less, 60 μm or less, 55 μm or less, 50 μm or less, 45 μm or less, 40 μm or less, 35 μm or less, 30 μm or less, 25 μm or less, 20 μm or less, 19 μm or less, 18 μm or less, 17 μm or less, 16 μm or less, 15 μm or less, 14 μm or less, 13 μm or less, 12 μm or less, 11 μm or less, 10 μm or less, 9.5 μm or less, 9 μm or less, 8.5 μm or less, 8 μm or less, 7.5 μm or less, 7 μm or less, 6.5 μm or less, 6 μm or less, 5.5 μm or less, 5 μm or less, 4.5 μm or less, 4 μm or less, 3.5 μm or less, 3 μm or less, 2.5 μm or less, 2 μm or less, 1.75 μm or less, 1.5 μm or less, 1.25 μm or less, 1 μm or less, 0.75 μm or less, or 0.5 μm or less). The thickness of the radioactive layer 112 can range from any of the minimum values described above to any of the maximum values described above. For example, the radioactive layer 112 can have a thickness of from 0.1 μm to 100 μm (e.g., from 0.1 μm to 50 μm, from 50 μm to 100 μm, from 0.1 μm to 20 μm, from 20 μm to 40 μm, from 40 μm to 60 μm, from 60 μm to 80 μm, from 80 μm to 100 μm, from 0.1 μm to 90 μm, from 0.1 μm to 80 μm, from 0.1 μm to 70 μm, from 0.1 μm to 60 μm, from 0.1 μm to 50 μm, from 0.1 μm to 40 μm, from 0.1 μm to 30 μm, from 0.1 μm to 20 μm, from 0.1 μm to 10 μm, from 0.1 μm to 5 μm, from 5 μm to 10 μm, from 0.1 μm to 2 μm, from 2 μm to 4 μm, from 4 μm to 6 μm, from 6 μm to 8 μm, from 8 μm to 10 μm, or from 0.5 μm to 9 μm). The devices can further comprise a first conducting layer 116 disposed above the plurality of spaced-apart three-dimensional elements 108, wherein the first conducting layer 116 is in electric contact with at least a portion of the plurality of spaced-apart three-dimensional elements 108. In some examples, the first conducting layer 116 forms a Schottky contact with at least a portion of the plurality of spaced-apart three-dimensional elements 108. In some examples, the first conducting layer 116 is in electric contact with each of the plurality of spaced-apart three-dimensional elements 108. In some examples, the first conducting layer 116 comprises a single continuous layer that is in electric contact with each of the plurality of spaced-apart three-dimensional elements 108. In some examples, the first conducting layer 116 comprises a plurality of conducting layers, wherein each of the plurality of conducting layers is in electric contact with one the plurality of spaced-apart three-dimensional elements 108, and each of the plurality of conducting layers is electrically connected to the other conducting layers, for example as shown in FIG. 8. The devices can further comprise a second conducting layer 118 disposed below the substrate 102, wherein the second conducting layer 118 is in electric contact with the bottom surface 106 of the substrate 102. The first conducting layer 116 and/or the second conducting layer 118 can comprise(s) a metal or a metal in combination with a transparent conducting oxide, a conducting polymer, a carbon material, or a combination thereof. In some examples, the first conducting layer 116, the second conducting layer 118, or a combination thereof comprise(s) a metal. The metal can comprise, for example, a metal selected from the group consisting of Al, Ti, Ni, Cu, Ga, Ag, Ir, Pt, Au, Cr, Mo, Pd, W, and combinations thereof. In some examples, the first conducting layer 116, the second conducting layer 118, or a combination thereof can comprise(s) a radioisotope. For example, the first conducting layer 116, the second conducting layer 118, or a combination thereof comprise(s) 63Ni. Examples of carbon materials include, but are not limited to, graphitic carbon and graphites, including pyrolytic graphite (e.g., highly ordered pyrolytic graphite (HOPG)) and isotropic graphite, amorphous carbon, carbon black, single- or multi-walled carbon nanotubes, buckminsterfullerene (C60), graphene, glassy carbon, diamond-like carbon (DLC) or doped DLC, such as boron-doped diamond, pyrolyzed photoresist films, and others known in the art. In some examples, the first conducting layer 116 and/or the second conducting layer 118 can comprise a transparent conducting oxide. In some examples, the first conducting layer 116 and/or the second conducting layer 118 can comprise a metal oxide. Examples of metal oxides include simple metal oxides (e.g., with a single metal element) and mixed metal oxides (e.g., with different metal elements). The metal oxide can, for example, comprise a metal selected from the group consisting of Cd, Cr, Cu, Ga, In, Ni, Sn, Ti, W, Zn, and combinations thereof. In some examples, the first conducting layer 116 and/or the second conducting layer 118 can comprise CdO, CdIn204, Cd2SnO4, Cr2O3, CuCrO2, CuO2, Ga2O3, In2O3, NiO, SnO2, TiO2, ZnGa2O4, ZnO, InZnO, InGaZnO, InGaO, ZnSnO, Zn2SnO4, CdSnO, WO3, or combinations thereof. In some examples, first conducting layer 116 and/or the second conducting layer 118 can comprise a conducting polymer such as polyacetylene, polyalanine, poly(3,4-ethylenedioxythiophene) polystyrene sulfonate, or combinations thereof. The devices can further comprise a first scintillation layer 120 disposed above the first conducting layer 116 and in radiative contact with the first conducting layer 116. As used herein, radiative contact means that the first scintillation later is disposed relative to the first conducting layer 116 such that a photon emitted from the first scintillation layer 120 can reach and penetrate the first conducting layer 116 and/or such that an energetic particle emitted from the radiation layer 112 (e.g., alpha particle, beta particle, etc.) can contact the first scintillation layer 120. In some examples, the devices can further comprise a second scintillation layer 122 disposed below the second conducting layer 118 and in radiative contact with the second conducting layer 118. In some examples, the first scintillation layer 120 can be in physical contact with the first conducting layer 116. In some examples, the second scintillation layer 122 can be in physical contact with the second conducting layer 118. The first scintillation layer 120, the second scintillation layer 122, or a combination thereof comprise(s) a scintillating material selected from the group consisting of a ZnS-based phosphor, NaI, CsI, cerium-doped lutetium yttrium orthosilicate (Ce:Lu1.8Y0.2SiO5, LYSO), bismuth germanate (Bi4Ge3O12, BGO), plastic, CeF3, europium doped calcium fluoride (Eu:CaF2), PbWO4, CdWO4, cerium doped cesium lithium yttrium chloride (Ce:Cs2LiYCl6, CLYC), Ce:LaCl3, CeBr3, Ce:LaBr3, praseodymium doped lutetium aluminum garnet (Pr:Lu3Al5O12, Pr:LuAG), cerium doped lutetium aluminum perovskite (Ce:LuAlO3, Ce:LuAP), Ce:Lu3Al5O12, cerium doped yttrium orthosilicate (Ce:Y2SiO5, Ce:YSO), cerium doped yttrium aluminum perovskite (Ce:YAlO3; Ce:YAP), cerium doped yttrium aluminum garnet (Ce:Y3Al5O12, Ce:YAG), and combinations thereof. In some examples, the first scintillation layer 120, the second scintillation layer 122, or a combination thereof can comprise(s) a scintillating material that can perform as an energy degrader. The first scintillation layer 120, the second scintillation layer 122, or a combination thereof can, for example, comprise a scintillating material that can be radioactive such that photons can be generated directly from within the scintillating materials upon exposure to radiation. For example, Na and I in Tl:NaI could be 22Na or 131I to form Tl:22NaI, Tl:Na131I, or Tl:22Na131I. 22Na undergoes beta+ decay, emitting positron and gamma rays, which can deposit energy in the plurality of spaced-apart three-dimensional structures and/or the substrate to generate charge or electricity and can also produce photons within Tl:NaI, subsequently depositing energy in the plurality of spaced-apart three-dimensional structures and/or the substrate to generate charge or electricity. Referring now to FIG. 9, in some examples the device 100 can further comprise a scintillating material 124 disposed within at least a portion of the plurality of coated cavities 114. In some examples, each of the plurality of coated cavities 114 can be uniformly filled with the scintillating material 124 relative to the other plurality of coated cavities 114. In some examples, each of the plurality of coated cavities 114 can be completely filled with the scintillating material 124, for example as shown in FIG. 10. In some examples, the plurality of coated cavities 114 can be overfilled with the scintillating material 124, for example as shown in FIG. 11. Referring now to FIG. 12, in some examples, the device 100 can further comprise a layer of a scintillating material 124 disposed on the radioactive layer 112, such that the plurality of coated cavities 114 are substantially coated by the layer of scintillating material 124. The scintillating material 124 can, for example, be selected from the group consisting of a ZnS-based phosphor, CsI, thallium-doped sodium iodide (Tl:NaI), cerium-doped lutetium yttrium orthosilicate (Ce:Lu1.8Y0.2SiO5, LYSO), bismuth germanate (Bi4Ge3O12, BGO), plastic, CeF3, europium doped calcium fluoride (Eu:CaF2), PbWO4, CdWO4, cerium doped cesium lithium yttrium chloride (Ce:Cs2LiYCl6, CLYC), Ce:LaCl3, CeBr3, Ce:LaBr3, praseodymium doped lutetium aluminum garnet (Pr:Lu3Al5O12, Pr:LuAG), cerium doped lutetium aluminum perovskite (Ce:LuAlO3, Ce:LuAP), Ce:Lu3Al5O12, cerium doped yttrium orthosilicate (Ce:Y2SiO5, Ce:YSO), cerium doped yttrium aluminum perovskite (Ce:YAlO3; Ce:YAP), cerium doped yttrium aluminum garnet (Ce:Y3Al5O12, Ce:YAG), and combinations thereof. In some examples, the scintillating material 124 can perform as an energy degrader. In some examples, the scintillating material 124 can comprise a radioactive material such that photons can be generated directly from within the scintillating material 124 upon exposure to radiation. For example, Na and I in Tl:NaI could be 22Na or 131I to form Tl:22NaI, Tl:Na131I, or Tl:22Na131I. 22Na undergoes beta+ decay, emitting positron and gamma rays, which can inject energy into the plurality of spaced-apart three-dimensional structures and/or the substrate to generate charge or electricity and can also produce photons within Tl:NaI, subsequently injecting energy into the plurality of spaced-apart three-dimensional structures and/or the substrate to generate charge or electricity. Referring now to FIG. 13, also disclosed herein are charge generating devices comprising: a substrate 102 having a top surface 104 and a bottom surface 106; a plurality of spaced-apart three-dimensional elements 108 disposed on the top surface 104 of the substrate 102; a plurality of cavities 110 formed by the plurality of spaced-apart three-dimensional elements 108, the plurality of cavities 110 being the area between the plurality of spaced-apart three-dimensional elements 108; a radioactive material 112 disposed within at least a portion of the plurality of cavities 110, thereby forming a plurality of at least partially filled cavities 110; a first conducting layer 116 disposed above the plurality of at least partially filled cavities 110 and the plurality of spaced-apart three-dimensional elements 108, wherein the first conducting layer 116 is in electric contact with the plurality of spaced-apart three-dimensional elements 108; a second conducting layer 118 disposed below the substrate 102, wherein the second conducting layer 118 is in electric contact with the bottom surface 106 of the substrate 102; a first scintillation layer 120 disposed above the first conducting layer 116 and in radiative contact with the first conducting layer 116; and a second scintillation layer 122 disposed below the second conducting layer 118 and in radiative contact with the second conducting layer 118. In some examples, the plurality of spaced-apart three-dimensional elements 108 are integrally formed with the substrate 102. In some examples, each of the plurality of cavities 110 can be uniformly filled with the radioactive material 112 relative to the other plurality of cavities 110. In some examples, each of the plurality of cavities 110 can be completely filled with the radioactive material 112, for example as shown in FIG. 14. In some examples, the plurality of cavities 110 can be overfilled with the radioactive material 112, for example as shown in FIG. 15. The radioactive material 112 can, for example, comprise a powder comprising a radioisotope. Referring now to FIG. 16, also disclosed herein are charge generating devices comprising: a substrate 102 having a top surface 104 and a bottom surface 106; a plurality of spaced-apart three-dimensional elements 108 disposed on the top surface 104 of the substrate 102; a plurality of cavities 110 formed by the plurality of spaced-apart three-dimensional elements 108, the plurality of cavities 110 being the area between the plurality of spaced-apart three-dimensional elements 108; a scintillating material disposed within at least a portion of the plurality of cavities 110, thereby forming a plurality of at least partially filled cavities 110; a first conducting layer 116 disposed above the plurality of at least partially filled cavities 110 and the plurality of spaced-apart three-dimensional elements 108, wherein the first conducting layer 116 is in electric contact with the plurality of spaced-apart three-dimensional elements 108; a second conducting layer 118 disposed below the substrate 102, wherein the second conducting layer 118 is in electric contact with the bottom surface 106 of the substrate 102; a first scintillation layer 120 disposed above the first conducting layer 116 and in radiative contact with the first conducting layer 116; and a second scintillation layer 122 disposed below the second conducting layer 118 and in radiative contact with the second conducting layer 118. In some examples, the plurality of spaced-apart three-dimensional elements 108 are integrally formed with the substrate 102. In some examples, each of the plurality of cavities 110 can be uniformly filled with the scintillating material 124 relative to the other plurality of cavities 110. In some examples, each of the plurality of cavities 110 can be completely filled with the scintillating material 124, for example as shown in FIG. 17. In some examples, the plurality of cavities 110 can be overfilled with the scintillating material 124, for example as shown in FIG. 18. Referring now to FIG. 19, in some examples, the scintillating material 124 is disposed as a layer on the plurality of spaced-apart three-dimensional elements 108, such that the first conducting layer 116 is in electric contact with the plurality of spaced-apart three-dimensional elements 108 through the layer of the scintillating material 124. In some examples, the device 100 can further comprise a radioactive layer 112 disposed on at least a portion of the plurality of spaced-apart three-dimensional elements 108 and the top surface 104 such that the plurality of cavities 110 and the top surface 104 are substantially coated by the radioactive layer 112, thereby forming a plurality of coated cavities 114 such that the scintillating material is disposed within at least a portion of the plurality of coated cavities 114, for example as described above and shown in FIG. 9-FIG. 11. The devices 100 can further include and/or be coupled to additional components. For example, the devices 100 can further comprise a charge storage device electrically coupled to the first conducting layer and the second conducting layer. In some examples, multiple devices 100 can be electrically coupled together. Methods of Making Also disclosed herein are methods of making the devices disclosed herein. For example, the methods can comprise forming the plurality of spaced-apart three-dimensional elements on the top surface of the substrate; and depositing the radioactive layer on at least a portion of the plurality of spaced-apart three-dimensional elements and the top surface such that the plurality of cavities and the top surface are substantially coated by the radioactive layer, thereby forming the plurality of coated cavities. Forming the plurality of spaced-apart three-dimensional elements on the top surface of the substrate can, for example, comprise a top-down or a bottom-up approach. In some examples, forming the plurality of spaced-apart three-dimensional elements can comprise a top down approach, such as etching a portion of the substrate to thereby form the plurality of cavities, and thus also forming the plurality of spaced-apart three-dimensional elements on the substrate where the spaced-apart three-dimensional elements are integrally formed with the substrate. Depositing the radioactive layer can, for example, comprise electrodeposition, printing, lithographic deposition, spin coating, drop-casting, zone casting, dip coating, blade coating, spraying, vacuum filtration, slot die coating, curtain coating, or combinations thereof. In some examples, wherein the device further comprises a scintillating material disposed within at least a portion of the plurality of coated cavities, the method further comprises depositing the scintillating material within at least a portion of the plurality of coated cavities. In some examples, therein the scintillating material comprises a radioactive scintillating material, the method further comprises depositing the radioactive scintillating material within at least a portion of the plurality of coated cavities. Depositing the scintillating material and/or the radioactive scintillating material can, for example, comprise electrodeposition, printing, lithographic deposition, spin coating, drop-casting, zone casting, dip coating, blade coating, spraying, vacuum filtration, slot die coating, curtain coating, or combinations thereof. A schematic diagram of an exemplary method of making a device is shown, for example, in FIG. 20. In some examples, the methods of making the device can comprise forming the plurality of spaced-apart three-dimensional elements on the top surface of the substrate; and depositing the scintillating material within at least a portion of the plurality of cavities. Methods of Use Also provided herein are methods of use of the devices described herein. The devices described herein can, for example, be used to convert radiation to electricity. The devices described herein can convert radiation to electricity with an efficiency of 10% or more (e.g., 15% or more, 20% or more, 25% or more, or 30% or more). In some examples, the devices described herein can convert radiation to electricity with an efficiency of from 25% to 35%. The devices can be used in various articles of manufacture, such as, for example, electronic devices, energy conversion devices, optical devices, optoelectronic devices, or combination thereof. Examples of articles of manufacture using the devices described herein can include, but are not limited to solar cells, fuel cells, photovoltaic cells, sensors, devices used in the Internet of Things (IoT), and combinations thereof. Such articles of manufacture can be fabricated by methods known in the art. The following examples are set forth below to illustrate the methods and results according to the disclosed subject matter. These examples are not intended to be inclusive of all aspects of the subject matter disclosed herein, but rather to illustrate representative methods and results. These examples are not intended to exclude equivalents and variations of the present invention which are apparent to one skilled in the art. Efforts have been made to ensure accuracy with respect to numbers (e.g., amounts, temperature, etc.) but some errors and deviations should be accounted for. Unless indicated otherwise, parts are parts by weight, temperature is in ° C. or is at ambient temperature, and pressure is at or near atmospheric. There are numerous variations and combinations of measurement conditions, e.g., component concentrations, temperatures, pressures and other measurement ranges and conditions that can be used to optimize the described process. Nuclear voltaic batteries are devices that use energy from the decay of radioactive isotopes to generate electricity. Nuclear batteries have a longer battery life compared to traditional chemical batteries. A schematic diagram of a nuclear battery is shown in FIG. 21. As shown in FIG. 21, when emitted particles (e.g., alpha particles, beta particles, photons) from the radioactive source interact with matter, the ionizing radiation crease electron-hole pairs in a semiconductor in the nuclear battery. The electron-hole pairs in the depletion region and near the depletion region can be separated by the built-in voltage of the nuclear battery. The separated electron-hole pairs can then diffuse and move to the electrodes to then be delivered to an external circuit. Because nuclear batteries have a longer battery life compared to traditional chemical batteries, they are attractive for many applications such as outer space exploration, in implantable medical devices (IMD) (e.g., deep brain neurostimulators, cochlear implants, gastric stimulators, cardiac defibrillators, pacemakers, foot drop implants, insulin pumps, etc.), and applications that involve harsh environment operation of the battery. For example, for outer space exploration applications, only a small percentage of sunlight reaches the outer perimeter of the solar system making solar power unfeasible. Furthermore, it is almost impossible to access the devices for outer space exploration once launched, therefor the device must function reliably over a long period of time without being accessed for recharge or maintenance. Similarly, for many implantable medical devices it can be difficult and/or dangerous to access the devices once implanted, therefor the device must function reliably over a long period of time without being accessed for recharge or maintenance. For both space exploration and implantable medical devices, it is further desirable for the power source to be small in size. Chemical batteries can malfunction when operated in an environment that is high in temperature, high in humidity, and/or high in radiation; nuclear batteries can be more stable under such harsh operating conditions. Traditional nuclear batteries currently suffer from a few challenges, such as low power conversion efficiency, small power output, and radiation damage to the semiconductor device. The wide band gap semiconductor materials used in traditional nuclear batteries theoretically have a small leakage current which can contribute to the low power conversion efficiency and low power output. A device with a large mean free path of the carrier can increase the number of electron-hole pairs that can be collected, and thus improve the performance of the nuclear battery. Further, a device using materials with a high atomic displacement energy, such as SiC, GaN, and diamond, can be resistant to radiation damage and can therefore improve the lifetime and efficiency of the device. Additionally, a device using materials which high thermal conductivity, such as SiC, can maintain stable working conditions even in a high temperature environment (optionally with extra cooling used). The X-ray response under voltaic mode of a commercial Si solar cell with an X-ray tube having a maximum power of 3.95 W as the radiation source is shown in FIG. 22, with a magnified view of further shown in FIG. 23. The X-ray tube power was adjusted by changing the voltage while the current was set to 79 μA. The fill factor can be calculated from the I-V curve according to: Fill Factor = P MAX P T = I MP × V MP I SC × V OC where PT, PMAX, VOC, and ISC are shown schematically in FIG. 23. A packaged device in a dual in-line chip carrier comprising a Schottky electrode and an Ohmic contact is shown in FIG. 24. The X-ray response under voltaic mode of a SiC Schottky diode is shown in FIG. 25. A thin scintillator layer positioned between the radiation source and the semiconductor in a nuclear battery, as shown in FIG. 26, can generate light for an improved power conversion efficiency. The stopping power for alpha particles of 5.49 MeV for Si and SiC is shown in FIG. 27. A thin phosphor layer (e.g., scintillator layer) can decrease the particle energy for a higher stopping power, matching up with the device length scale (e.g., ˜1 μm). Without the scintillator, the device relies on direct detection, whereas with the scintillator the device can use combined direct and indirect detection, as shown schematically in FIG. 28. The device with the scintillator layer can, for example, further include a metal layer between the scintillator and semiconductor which can block the photons from the scintillator, as shown schematically in FIG. 29. The power output for devices shown schematically in FIG. 29 comprising a solar cell and a thin layer of ZnS(Ag) as a scintillator or comprising a solar cell, an aluminum foil layer, and a thin layer of ZnS(Ag) is shown in FIG. 30. Compared to the devices without a scintillator, the power output increased 787% by adding ZnS(Ag). The power output for devices shown schematically in FIG. 29 comprising a solar cell, an aluminum foil layer, and a thin layer of CLYC (cerium doped cesium lithium yttrium chloride (Ce:Cs2LiYCl6) as the scintillator; a solar cell, a layer of optical grease, and a thin layer of CLYC (cerium doped cesium lithium yttrium chloride (Ce:Cs2LiYCl6) as the scintillator; and a solar cell, a layer of optical grease, a thin layer of CLYC (cerium doped cesium lithium yttrium chloride (Ce:Cs2LiYCl6) as the scintillator; and a Teflon reflector is shown in FIG. 31. The power output increase ˜58% by adding CLYC. The power output increased an additional ˜1% after adding a Teflon reflector. The photon emission spectra properties of various scintillators are described below in Table 1. When compared to the External Quantum Efficiency (EQE) of a Si solar cell and SiC, it can be seen that the photons emitted from ZnS(Ag) match well with a Si solar cell, but none of the three phosphors described in Table 1 match well with SiC since the emission photons of the scintillators are too low in energy to include a noticeable power increase. Deep UV emission scintillators with high light yield would need to be applied to SiC to enhance the efficiency. TABLE 1Photon emission properties of ZnS(Ag),CLYC, and polyvinyltoluene (PVT).ZnS(Ag)CLYCPVTPeak emission wavelength (nm)450370435Corresponding energy (eV)2.753.352.85Scintillator light yield (Photons/MeV)58,50020,0009,000 An alpha voltaic battery based on a Si semiconductor device was designed and fabricated. An alpha voltaic device potentially has a higher output power as alpha particles have higher relative energy than beta particles. Further, an alpha voltaic device can be smaller in size and easier to shield. However, the higher energy alpha particles will degrade materials faster than the lower energy beta particles. A customized 4 μCi Am-241 source was used as the alpha source in the device (diameter 17.9 mm, FIG. 32). Further, PVT was coated on the alpha source as a scintillator and energy degrader (FIG. 33). An EMCCD camera was used for emission photon imaging of the PVT-coated Am-241 source to determine if there were any alpha induced photons (FIG. 34-FIG. 35). Additionally, 2D and 3D Si p-n diodes were investigated. Images of an unetched device (2D) are shown in FIG. 36 and FIG. 37. The 2D Si device was partially covered by the Am-241 source (FIG. 38). The 2D Si p-n diode response to the Am-241 source was tested (FIG. 39). When the Am-241 source was applied to the unetched 2D Si device, the device's maximum power output at atmosphere was about 3.3 pW and at 30 mTorr was about 3.59 pW with about 70% of the area of the 2D Si device covered by the Am-241 source. A 3D etched device has a larger effect area than a 2D unetched device, which can theoretically yield higher efficiency. A 3D device array test set up was designed to utilize as much of the source area as possible (FIG. 40). During the test, the set up (FIG. 40) remained the same, but a different number of devices were connected to investigate the power generation (FIG. 41). A discussed above, traditional nuclear batteries can suffer from a few challenges, such as low power conversion efficiency, small power output, and radiation damage to the semiconductor device. The devices described herein address these challenges through the use of a scintillator, 3D structures, and wide band-gap materials. Other advantages which are obvious and which are inherent to the invention will be evident to one skilled in the art. It will be understood that certain features and sub-combinations are of utility and may be employed without reference to other features and sub-combinations. This is contemplated by and is within the scope of the claims. Since many possible embodiments may be made of the invention without departing from the scope thereof, it is to be understood that all matter herein set forth or shown in the accompanying drawings is to be interpreted as illustrative and not in a limiting sense. The devices and methods of the appended claims are not limited in scope by the specific devices and methods described herein, which are intended as illustrations of a few aspects of the claims. Any devices and methods that are functionally equivalent are intended to fall within the scope of the claims. Various modifications of the devices and methods in addition to those shown and described herein are intended to fall within the scope of the appended claims. Further, while only certain representative devices and methods disclosed herein are specifically described, other devices and methods also are intended to fall within the scope of the appended claims, even if not specifically recited. Thus, a combination of steps, elements, components, or constituents may be explicitly mentioned herein or less, however, other combinations of steps, elements, components, and constituents are included, even though not explicitly stated. The term “comprising” and variations thereof as used herein is used synonymously with the term “including” and variations thereof and are open, non-limiting terms. Although the terms “comprising” and “including” have been used herein to describe various embodiments, the terms “consisting essentially of” and “consisting of” can be used in place of “comprising” and “including” to provide for more specific embodiments of the invention and are also disclosed. Other than where noted, all numbers expressing geometries, dimensions, and so forth used in the specification and claims are to be understood at the very least, and not as an attempt to limit the application of the doctrine of equivalents to the scope of the claims, to be construed in light of the number of significant digits and ordinary rounding approaches. As used in the description and the appended claims, the singular forms “a,” “an,” and “the” include plural referents unless the context clearly dictates otherwise. Thus, for example, reference to “a composition” includes mixtures of two or more such compositions, reference to “an agent” includes mixtures of two or more such agents, reference to “the component” includes mixtures of two or more such components, and the like. “Optional” or “optionally” means that the subsequently described event or circumstance can or cannot occur, and that the description includes instances where the event or circumstance occurs and instances where it does not. It is understood that throughout this specification the identifiers “first” and “second” are used solely to aid in distinguishing the various components and steps of the disclosed subject matter. The identifiers “first” and “second” are not intended to imply any particular order, amount, preference, or importance to the components or steps modified by these terms. The term “or combinations thereof” as used herein refers to all permutations and combinations of the listed items preceding the term. For example, “A, B, C, or combinations thereof” is intended to include at least one of: A, B, C, AB, AC, BC, or ABC, and if order is important in a particular context, also BA, CA, CB, CBA, BCA, ACB, BAC, or CAB. Continuing with this example, expressly included are combinations that contain repeats of one or more item or term, such as BB, AAA, AB, BBC, AAABCCCC, CBBAAA, CABABB, and so forth. The skilled artisan will understand that typically there is no limit on the number of items or terms in any combination, unless otherwise apparent from the context. Unless defined otherwise, all technical and scientific terms used herein have the same meanings as commonly understood by one of skill in the art to which the disclosed invention belongs. Publications cited herein and the materials for which they are cited are specifically incorporated by reference. |
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048209295 | abstract | A dynamic infrared simulation cell comprising a photoconductive layer, a first conductive layer, a second conductive layer, and an external energy source. The first conductive layer is affixed about one side of the photoconductive layer and is transmissive with respect to radiation of known energy. The second conductive layer is affixed about the other side of the photoconductive layer. The external energy source connects to the first conductive layer and the second conductive layer. The photoconductive layer is a plurality of segments of photoconductive silicon material. A dielectric material fills the areas between the segments of photoconductive material. The first conductive layer is a transparent layer of gold material. The second conductive layer is a layer of aluminum material. An enclosure is available for maintaining the cell in a darkened environment. |
claims | 1. A multipurpose device for transmitting radiation from a source to an object, comprising two reflectors each shaped as a truncated segment of a curved surface and both arranged to produce a combined focal zone, a distributed radiation source positioned in the aperture plane of at least one of the reflectors or in one of the focal areas of each reflector, and an object placed in the combined focal area of both reflectors. 2. A multipurpose device as claimed in claim 1, wherein the curved surface of each reflector is a spherical or cylindrical surface, and the distributed radiation source is placed in the aperture plane of at least one of the reflectors. 3. A multipurpose device as claimed in claim 1, wherein the reflectors are arranged opposite one another to produce a combined focal area. 4. A multipurpose device as claimed in claim 1, wherein the reflectors are arranged at an angle to one another to produce a combined focal area. 5. A multipurpose device as claimed in claim 1, further provided with at least one pair of spherical or cylindrical reflectors in a plane normal to the plane of the first pair of antenna reflectors. 6. A multipurpose device as claimed in claim 1, wherein the curved surface of the reflectors comprises at least a pair of elliptical cylinders positioned opposite one another to produce a single combined focal area for an object to be placed therein, distributed sources being arranged in the two other focal areas. 7. A multipurpose device as claimed in claim 5, wherein one elliptical reflector only is used in at least two planes, one distributed source being placed in the focal area of each reflector, and an object being placed in the combined focal area of said reflectors. |
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description | Referring first to FIG. 1, the aircraft nuclear propulsion plant includes a compressor section 20, a reactor section 21, a turbine section 22, and ah exhaust nozzle 23. Extending between compressor section 20 and turbine section 22 and passing through the center of reactor section 21 is a shaft 24 coupling the compressor and turbine sections. Reactor section 21 includes a nuclear reactor 25 surrounded by a shield 26 while shield plugs 27 and 28 are disposed at opposite ends of the reactor. FIG. 1 also shows combustion cans 29 which will be described in more detail hereinafter. FIG. 2 discloses reactor 25 in somewhat more detail. Reactor 25 comprises an annular core 30 which is enclosed within and spaced from a pressure vessel 31 leaving an annular channel 32 around the periphery of the pressure vessel 31. Twenty-four combustion cans or ducts 29 are disposed in annular channel 32. Reactor core 30 comprises an annular active portion 33, an annular inner reflector 34, an annular outer reflector 35, radial coolant outlet channels 36 and radial coolant inlet channels 37. Active portion 33 includes rectangular fuel zones 38 and wedge-shaped moderator zones 39. It will be noted that coolant outlet channels 36 communicate with the interior of combustion cans 29 while coolant inlet channels 37 communicate with annular channel 32. Pressure vessel 31 is provided with a horizontal flange 39a permitting radial removal of individual fuel segments. Reactor 25 will now be described in detail by referring to FIGS. 3 and 4. As mentioned previously, combustion cans 29 are located around the periphery of reactor 25 in annular channel 32. Combustion cans 29 have a dual functionxe2x80x94they serve as burners for an auxiliary chemical combustion system and as outlet air ducts for the reactor 25. The chemical combustion system will only be described cursorily because it forms no part of the present invention. Located in combustion cans 29 are inner liners 40 which extend from the inlet end thereof to a point just short of the midpoint of the combustion cans. A valve 41 is located at the inlet end of combustion cans 29 and a swirl cap 42 is located just inside inner liner 40. Cross-fire tubes 43 connect adjacent combustion cans 29. Baffles 44 are employed to direct air from outlet channels 36 into combustion cans 29 with a swirling motion to promote mixing. Combustion cans 29 are circular in cross section at the inlet end thereof and gradually change to rectangular at the outlet end thereof. The cross-sectional area between cans decreases linearly throughout the length of the core 30 so that the air introduced into the core is at a constant speed and pressure across the length of the core. The annular volume of the reactor is formed by the radial assembly of 24 identical segments 45, each of which includes a part of end reflectors 46 as well as a part of outer reflector 35 and a part of active portion 33. Segments 45 are defined by side plates 47 and end plates 48. Side plates 47 cover the outer reflector 35 and end reflector 46 but not the active portion 33. L-shaped structural beams 49 are fastened to the top of plates 47 and extend the full length of the segment 45. Beams 49 serve as primary structural support for the reactor core 30, supporting segments 45 from shield plugs 27 and 28. Side plates 47 of adjacent segments 45 are spaced to form outlet channels 36. Dividers 50 separate the end reflectors 46 from the outer reflector 35 and active portion 33. Extending between side plates 47 at spaced intervals between dividers 50 are bulkheads 51 which divide the outer reflector 35 into a plurality of sections 52. The two outermost bulkheads 51A are solid, but the remaining bulkheads 51 have a central radial elongated slot 53 therein (see FIG. 7). The distance between end plates 50 and the outermost bulkheads 51A is only half that between bulkheads 51. Each section 52 of outer reflector 35 is comprised of a pair of rows of spaced aligned flat slabs 54 of beryllium extending between side plates 47 and fastened thereto by means of bolts 55 passing through flanges 56 on slabs 54. Tie rods 57 extend through flanges 56, while tie rods 58 extend through slabs 54 to hold them securely together. As shown in FIG. 4, the pairs of slabs 54 are separated to form coolant inlet channels 37. There are 13 slabs 54 of beryllium in each row of slabs in full-size sections 52 and only 7 in each of half-size sections at the two ends of the outer reflector 35. Active portion 33 comprises a plurality of abutting radially arranged sections 59 each having a pair of end plates 60 having a depressed portion 61 at the center thereof (see FIG. 5), a bottom plate 62, a pair of fuel elements 63 and a pair of moderator elements 64. Seven full-size and two one-half-size fuel elements 63 are aligned end to end to form a fuel zone 38 and seven full-size and two one-half size-moderator elements are aligned end to end to form a moderator zone 39. Fuel elements 63 are rectangular in form and include a frame 65 and a plurality of interlaced wires 66 extending across the frame. Wires 66 are formed of a suitable refractory metal or ceramic material such as stainless steel, a xe2x80x9cNichromexe2x80x9d alloy, iron-chromium-yttrium alloys, chromium-titanium alloys, clad graphite, aluminum oxide, and thorium oxide and include a fully enriched uranium dioxide core. xe2x80x9cNichromexe2x80x9d alloys include an alloy consisting of 15-16% chromium, 59-62% nickel, about 24% iron, and 0.1% carbon and an alloy of 80% chromiumxe2x80x9420% nickel. Fuel elements 63 are 4xe2x80x3xc3x976xe2x80x3xc3x970.75xe2x80x3. Moderator zone 39 is comprised of moderator elements 64 located on the inner face of each of the fuel elements 63. Moderator elements 64 consist of a row of wedge-shaped, radially-arranged bars 67 of metallic hydride, such as yttrium hydride, or of zirconiumhydride having a NH of 6, where NH is defined as the number of hydrogen atoms per cubic centimeterxc3x9710xe2x88x9222 at room temperature, which are tied together by rods 68 which penetrate depressed portions 61 of end plates 60. Moderator elements 64 are spaced to form coolant inlet channels 37. The cleft running downthrough the moderator forming an extension of inlet passage 37 is for the purpose of minimizing friction in the incoming channel and conducting cool air to the lower end of the moderator pieces to prevent overheating. Active portion 33 is supported from outer reflector 35 by a quick release latch 69. End reflectors 46 are composed of three pie-shaped slabs 70 of beryllium totalling 8xe2x80x3 in thickness. A scoop 71 is located in each segment 45 at the corner of end reflector 46 nearest the inlet end of combustion cans 29. Scoop 71 directs air into a tapering inlet manifold 72 located between front end plate 48 and front end reflector 46. A tapering collecting manifold 73 is located between front end reflector 46 and divider 50 which includes openings 74 about the periphery thereof. A cylindrical cover plate 74A prevents direct communication between channel 32 and end sections 52. At the aft end of the core another scoop 75 is located on the periphery of rear end plate 48 and serves to direct air from a channel 75A into a tapering inlet manifold 76 located between rear end reflector 46 and rear end plate 48. A tapering collecting manifold 77 is located between rear end reflector 46 and divider 50. Openings 74 are also included in divider 50 and cover plate 77A prevents short circuiting of the coolant air. The reactor is controlled by control blades 78 which contain notches 79 in the lower edge thereof. Control blades 78 are disposed in coolant inlet channels 37 passing through slots 53 in bulkheads 51. Roller bearings 79A guide blades 78 between solid bulkheads 51A. Control blades 78 are operated by a hydraulic piston actuator (not shown) located just outside the pressure shell, which drives a push rod 80. A linkage 81 converts the axial motion of the blade 78 into a 4xc2xdxe2x80x3 radial stroke within reflector 35, the notches 79 corresponding to the location of bulkheads 51. A spring (not shown) serves to hold the blades inserted when the actuator is removed and causes the system to be fail-safe in the event of actuator failure. To make the fueled Nichrome wire, a tube of xe2x80x9cNichromexe2x80x9d 9.5xe2x80x3 long, 0.70xe2x80x3 in diameter and about 0.160xe2x80x3 wall thickness is obtained. A blend of 70% xe2x80x9cNichromexe2x80x9d and 30% UO2 powder, for example, is green-pressed, sintered and coined to 93% theoretical density. The compacts are then assembled into the tube with the ends of the tube capped with NiCr plugs. The compacts contained in the tube are subjected to 10 tons pressure to secure the compacts in place. After a diffusion treatment of two hours at 2100xc2x0 F. in a hydrogen atmosphere, the ends are welded closed. At this point the tube is rod-rolled (hot) to a diameter of 0.275xe2x80x3. It is then cold-drawn by successive steps to the desired diameter of 35 mils. The coils of wire are annealed prior to every drawing operation. With a target of 4 mils for the clad thickness, the usual results range from 3-7 mils. The bond between the clad and the fueled core is generally sound. To fabricate the fuel elements the finished fueled wire is straightened to remove curl. Then 60 to 80 wires about five feet long each are stretched in a grooved metal plate so they are in a plane parallel to each other and precisely 25 mils apart. The wires are then heavily coated with an epoxy resin containing a small amount of braze powder. When the resin has hardened, the wires plus the resin form a strip which is then cut into squares corresponding to the size of the fuel element desired. The cut strips are then stacked in a criss-cross fashion to the desired fuel element thickness, usually about xc2xexe2x80x3. Alignment is made positive by placing the stack in a jig which prevents the wires from moving during subsequent operations. Then the stacked wires, together with the jig, are inserted in a brazing furnace under a moderate clamping pressure. As the temperature rises to several hundred degrees F., the resin melts and runs off, leaving most of the braze material behind on the wires. As the brazing temperature of 2150xc2x0 F. is reached, the joints where the wires cross each other are brazed together. It has been found that some joints are not brazed, but this does not significantly weaken the brazed lock since the fraction of missed joints is low. The zirconium hydride bars are made by surrounding a zirconium bar with a hydrogen atmosphere, at an elevated temperature, allowing the system to come to equilibrium, and then cooling under controlled temperature-pressure relationship as described in patent application Ser. No. 785,542 filed Jan. 7, 1959 on behalf of James B. Vetrano. In operation, air discharged from compressor section 20 flows straight aft through annular passage 32 and is distributed radially inward along the full length of the reactor. The flow enters coolant inlet channels 37, passes first through reflector slabs 54, then enters active portion 33 and turns laterally to flow first through moderator elements 64 and then through fuel elements 63 and is then discharged outward radially through coolant channels 36 into the side of chemical combustion cans 29. The flow then proceeds aft through the turbine and exhaust nozzle of the power plant. A small portion of the air from the compressor is picked up by scoops 71 and 75 respectively which direct it into manifolds 72 and 76 respectively. The air cools end reflectors 46 by passing through perforations (not shown) therein and is collected in manifolds 73 and 77. From manifolds 73 and 77 the air proceeds through openings 74 to make a right-angle turn and enter inlet channels 37 in half-size sections 52. In half-size sections 52 the air travels the same path as in the remainder of the core. Thus the air employed to cool the end reflectors serves a dual function as it also cools a half-size fuel element. The high performance of this reactor results from the effectiveness of the heat transfer. The very energetic heat transfer that takes place between the fuel element wires and the air is caused by the large heat transfer coefficient together with the large surface area of the wires. The heat transfer coefficient is known to be 267 B/hr. ft.2xc2x0 F. at cruise and 450 at emergency power. The magnitude of the coefficient is attributable to the interrupted path and turbulence of the air. The fuel element contains 630 square feet of surface area for every cubic foot of matrix, yielding a total heat transfer area of 2520 square feet. The total frontal area of the fuel elements, 64 square feet, is so great that the air trickles through them at a mean velocity of about 50 feet per second. One of the best features of this reactor is the nearly isothermal condition achieved in the wires because of the steep nuclear power gradient existing through the xc2xe-inch thickness of the fuel element. This occurs because the fission-producing neutrons emanating from the moderator impinge with full intensity upon the front wires but fall off to about one-fifth of the intensity by the time that they reach the rear wires. This attenuation is caused by their passage through the fuel. The result is that all wires are worked at nearly their full heat transfer capacity, the front wires running at 1417xc2x0 F. and the rear wires at 1968xc2x0 F. while producing air at 1850xc2x0 F. The compactness of the reactor results from use of the reversed folded-flow principle. Folded flow alone reduces the core volume by only about 10% over that of a straight-through reactor. Folded flow combined with flow reversal reduces the core volume by 50%, because the distance any given air element must travel through the core is reduced to a minimum. This reduction in core volume is highly desirable, of course, because of the concomitant reduction in the amount of shielding required. One advantage of the structure described arises from the relative location of the moderator and the fissionable material. It will be noted that the entire air flow passes through the moderator before it passes through the fuel. Thus, the moderator is bathed in relatively cool air which holds the temperature of the moderator down. Thus, even though the temperature of the fuel is relatively high, the temperature of the moderator is relatively low. This is important because unclad zirconium hydride cannot be employed at a temperature much above 1200xc2x0 F. Another advantage is that virtually all components of the reactor are bathed in cool inlet air. The relatively cool inlet air from the compressor is directed over the outer reflector, the end reflector, the moderator, and part of the core structure. In addition the location of the control blades is such that they are bathed in cool inlet air. Also a small proportion of the air from the compressor is directed over the inner reflector by means not shown. It will be understood that the invention is not to be limited by the details given herein but that it may be modified within the scope of the appended claims. |
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claims | 1. A fuel bundle for a nuclear reactor comprising:a first fuel element including thorium dioxide;a second fuel element including uranium having a first fissile content; anda third fuel element including uranium having a second fissile content different from the first fissile content,wherein the first fuel element, the second fuel element, and the third fuel element are arranged such that a common coolant flows over each of the first fuel element, the second fuel element, and the third fuel element. 2. The fuel bundle of claim 1, wherein the uranium having the first fissile content includes recycled uranium having a fissile content of approximately 0.72 wt % of 235U to approximately 1.2 wt % of 235U. 3. The fuel bundle of claim 2, wherein the uranium having the second fissile content includes recycled uranium having a fissile content of approximately 0.72 wt % of 235U to approximately 1.2 wt % of 235U. 4. The fuel bundle of claim 2, wherein the uranium having the second fissile content includes slightly enriched uranium having a fissile content of approximately 0.9 wt % of 235U to approximately 3 wt % of 235U. 5. The fuel bundle of claim 2, wherein the uranium having the second fissile content includes natural uranium having a fissile content of approximately 0.71 wt % of 235U. 6. The fuel bundle of claim 2, wherein the uranium having the second fissile content includes low enriched uranium having a fissile content of approximately 3 wt % of 235U to approximately 20 wt % of 235U. 7. The fuel bundle of claim 1, wherein the uranium having the first fissile content includes slightly enriched uranium having a fissile content of approximately 0.9 wt % of 235U to approximately 3 wt % of 235U. 8. The fuel bundle of claim 7, wherein the uranium having the second fissile content includes slightly enriched uranium having a fissile content of approximately 0.9 wt % of 235U to approximately 3 wt % of 235U. 9. The fuel bundle of claim 7, wherein the uranium having the second fissile content includes natural uranium having a fissile content of approximately 0.71 wt % of 235U. 10. The fuel bundle of claim 7, wherein the uranium having the second fissile content includes low enriched uranium having a fissile content of approximately 3 wt % of 235U to approximately 20 wt % of 235U. 11. The fuel bundle of claim 1, wherein the uranium having the first fissile content includes natural uranium having a fissile content of approximately 0.71 wt % of 235U. 12. The fuel bundle of claim 11, wherein the uranium having the second fissile content includes low enriched uranium having a fissile content of approximately 3 wt % of 235U to approximately 20 wt % of 235U. 13. The fuel bundle of claim 1, wherein the uranium having the first fissile content includes low enriched uranium having a fissile content of approximately 3 wt % of 235U to approximately 20 wt % of 235U. 14. The fuel bundle of claim 13, wherein the uranium having the second fissile content includes low enriched uranium having a fissile content of approximately 3 wt % of 235U to approximately 20 wt % of 235U. 15. The fuel bundle of claim 2, wherein the uranium included in at least one of the second fuel element and the third fuel element is included with at least one of recycled uranium having a fissile content of approximately 0.72 wt % of 235U to approximately 1.2 wt % of 235U, depleted uranium having a fissile content of approximately 0.2 wt % of 235U to approximately 0.5 wt % of 235U, slightly enriched uranium having a fissile content of approximately 0.9 wt % of 235U to approximately 3 wt % of 235U, natural uranium having a fissile content of approximately 0.71 wt % of 235U, and low enriched uranium having a fissile content of approximately 3 wt % of 235U to approximately 20 wt % of 235U. 16. The fuel bundle of claim 2, wherein the uranium included in at least one of second fuel element and the third fuel element is included with a burnable poison. 17. The fuel bundle of claim 2, wherein the thorium dioxide included in the first fuel element is included with a burnable poison. 18. The fuel bundle of claim 1, wherein the first fuel element includes a rod of thorium dioxide. 19. The fuel bundle of claim 1, wherein the second fuel element includes a rod of the uranium having the first fissile content. 20. The fuel bundle of claim 1, wherein the third fuel element includes a rod of the uranium having the second fissile content. 21. The fuel bundle of claim 1, wherein the first fuel element includes a tube containing the thorium dioxide. 22. The fuel bundle of claim 1, wherein the second fuel element includes a tube containing the uranium having the first fissile content. 23. The fuel bundle of claim 1, wherein the third fuel element includes a tube containing the uranium having the second fissile content. 24. The fuel bundle of claim 1, wherein the second fissile content is higher than the first fissile content. 25. A nuclear reactor comprising: a tube of pressurized fluid; and the fuel bundle of claim 1. |
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052308596 | abstract | To ensure safe removal of flammable explosive gaseous mixtures, the flammable mixtures are burned and/or recombined in partial volumes of the container separated from the rest of the container by metal grilles (18,19). The combustion is thereby prevented from spreading by the principle of the Davy safety lamp. The combustion and/or recombination in the partial volumes is effected by means of ignition sources (20), such as electric sparks, hot surfaces, open flames and/or catalytic surfaces. The heat energy released during combustion is transferred by means of cooling devices such as heat pipes. The temperature of the grille is sued to control the energy supply of the ignition source(s) and should never exceed approximately 2/3 of the ignition temperature of the flammable gaseous mixture; when the temperature reaches 3/4 of the ignition temperature, a fuse cuts off the energy supply. If the grille suffers mechanical damage, it touches an electrically insulated internal grille. This spark-free contact triggers an electrical switch which cuts off the power supply to the ignition and/or glow plugs. Detonatable H2-air mixtures can thereby be removed without the risk of explosion. |
abstract | To reduce X-ray exposure, an area of interest is selected in the image. The image of the selected area is updated frequently, comparable to rate of updates used today for the whole image. The rest of the image is updated at a significantly lower rate. Since the area of interest normally is a small part of the overall area, the total exposure is reduced significantly. A movable X-ray shield placed near the X-ray source blocks the radiation from areas outside the area of interest. The shield automatically retracts when the complete image is updated. The area of interest can be selected by the user or automatically selected based on activity in the image. |
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051715198 | description | DETAILED DESCRIPTION OF THE INVENTION Certain aspects of the use of a full system chemical decontamination process to clean the primary reactor fluid system of radioactive "crud" is disclosed in co-pending Application Ser. No. 07/621,120, filed Nov. 26, 1990, now U.S. Pat. No. 5,089,216, entitled "System for Chemical Nuclear Reactor Primary Systems," Application Ser. No. 07/621,129, filed Nov. 26, 1990, now U.S. Pat. No. 5,089,217, entitled "Clean-up Subsystem for Chemical Decontamination of Nuclear Reactor Primary System," and Application Ser. No. 07/621,130, filed Nov. 26, 1990, entitled "Resin Processing System," all of which are incorporated herein by reference. The designing of such a system to allow for the installation of the decontamination equipment outside of the containment chamber is necessary for certain nuclear plants. The containment chamber is a heavily shielded area which encompasses the critical nuclear reactor equipment which handles radioactive materials, such as the reactor vessel, the primary process fluid system and the steam generation equipment. An "outside of containment" configuration has several advantages over an "inside-of-containment" configuration. The "inside-of-containment" configuration is defined as having all of the decontamination equipment which handles radioactive material inside of the containment chamber. First, all of the decontamination equipment set-up and shielding fabrication is performed off critical path. This means that the reactor does not have to be shut-down while the decontamination equipment is being installed or disassembled, thus reducing costs. Second, the spent resin handling operations are performed off critical path and thus costs are reduced. Third, equipment size and shielding are not limited by the size of the equipment hatch, thus allowing for a more flexible design. It has been found that the optimum interface between the chemical decontamination system and the primary reactor coolant system is via the residual heat removal (RHR) system. This is further detailed in a co-pending application Ser. No. 07/621,120, filed Nov. 26, 1990, now U.S. Pat. No. 5,089,216, entitled "System for Chemical Decontamination of Nuclear Reactor Primary Systems," and incorporated herein by reference. The connection between an outside of containment decontamination system and the primary reactor coolant system is preferably made at a point outside of the containment chamber. In this way, no new lines must be established through the containment chamber. Most plants contain a part of the RHR system outside of the containment chamber, or a line from the RHR system coming out of the containment chamber, and therefore that system is an ideal system to gain access to the primary reactor coolant system. In the preferred embodiment, the connection between the decontamination system inlet and the decontamination outlet are on the same RHR system line. The inlet decontamination connection is made upstream from the outlet decontamination connection. Placed between the inlet connection and the outlet connection is a flow regulation valve which directs the primary process fluid from the RHR system to the decontamination system. If a connection to the primary reactor coolant system cannot be made effectively outside of the containment chamber, then new lines must be established between the decontamination system and the primary reactor coolant system through the containment wall. One preferred access way would be to use the equipment hatch as a piping route. In this way, when the decontamination is complete, the piping can be removed and thus a permanent new line penetration through containment is avoided. In this connection design, the decontamination system connection must be made on the critical path and therefore during a reactor shut-down. The process piping between the RHR system and the decontamination system may be long, thus requiring precautions against the unlikely event of leakage. Therefore, the piping is preferably provided with safety features. These safety features may include a secondary envelope medium around the piping lines, such as double wall piping, trenches, curbs, sumps, or encasement in concrete. Certain aspects of the equipment utilized in the full system chemical decontamination process and the flow connections between these systems are set forth in copending application Ser. No. 07;621,129 filed Nov. 26, 1990, now U.S. Pat. No. 5,089,217, entitled "Clean Up Subsystem for Chemical Decontamination of Nuclear Reactor Primary Systems," and Application Ser. No. 07/621,130, filed Nov. 26, 1990 entitled "Resin Processing System," and both of these are incorporated herein by reference. However, since the entire decontamination system is not discussed in either of those two applications, the entire system will be set forth hereinafter. The preferred embodiment of the entire system decontamination process as described hereinafter is sized to operate within a "four-loop" reactor. Such a reactor has four reactor coolant system steam generation units. The same process may be employed for smaller plants utilizing "two-loop" and "three-loop" designs. The smaller plant designs would employ a smaller equipment design which would be roughly proportional to the reduction in wetted surface area compared to the "four-loop" design. Referring now to FIG. 1, process fluids from the primary system are sent via line 57 to an optional back-flush filter system 701. The back-flush filter system 701 is provided to filter suspended solids found in the primary system which are removed from the primary system during the decontamination. This system also removes manganese dioxide colloids or particles which are generated during the known CAN-DEREM and LOMI techniques. Certain chemical decontamination processes may not generate suspended solids in the primary process fluids. Therefore, the utilization of the back-flush filter system 701 is considered to be optional. The process fluids first enter a back-flush filter 301. This back-flush filter 301 can be periodically back-flushed by use of accumulator 300 which has an inlet nitrogen line 40 connected thereto. Also, line 41 provides demineralized water to aid in the back-flushing of the back-flush filter 301. When a back-flush step is in process the back-flushed material will be collected into filtrate collection tank 101 and can be pumped via filtrate transfer pump 201 along line 42 to any of the spent resin storage tanks 121 shown on FIG. 4. After exiting the back-flush filter 301 the process fluids enter the post filters 302. The back-flush filter 301 is preferably sized to handle the full flow rate of the decontamination system; in this case about 1500 gallons per minute (gpm), and has a 5-20 micron filter. The accumulator 300 is preferably sized to have a capacity sufficient to perform a complete backflush, in this case about 30 gallons. The filtrate collection tank 101 is preferably sized to have a capacity of several backflush volumes; arbitrarily set at about 400 gallons and the filtrate collection pump 201 is preferably sized to empty the filtrate collection tank 101 in less than an hour with a capacity of about 10-50 gpm. The post filters 302 are designed to operate in parallel and preferably have a combined flow rate equal to the backflush filter 301 of about 1500 gpm. Most preferably, four such filters are employed, each having a capacity of about 375 gpm and having about a 1 micron filter rating. Any type of filtering equipment may be used to constitute the backflush filter system 701. This system is not limited to a back-flush filter and other embodiments may include the use of cartridge filters with or without the use of a preliminary back-flush filter. After exiting the post filters 302 the process fluid travels through line 43, and now referring to FIG. 2, enters the demineralizer system 702. The demineralizer system 702 is provided to remove the chemicals which are added during the chemical decontamination process and radioactive corrosion products. This system is shown as being comprised of demineralizers 80, 81, and 82. Preferably, at least two demineralizers are employed, however, more than two demineralizers may be used. In the preferred embodiment, demineralizer 82 is a Regen demineralizer and has a total volume of about 400 ft..sup.3 (11.3 m.sup.3). Most preferably this Regen demineralizer contains three demineralizer vessels 611 which each have a volume of about 133 ft..sup.3 (3.8 m.sup.3). In the preferred embodiment two other demineralizers 80, 81 are also employed and each have a volume capacity of 600 ft.sup.3 (17.0 m.sup.3). Most preferably the demineralizers 80, 81 each contain three demineralizer vessels 611 which each have a volume of about 200 ft..sup.3 (5.7 m.sup.3). The amount of ion exchange resin used for decontamination is determined by the amount of deposits which have been produced in the RCS system. A small amount of deposits in a RCS would require less resin than that required for a heavily contaminated RCS facility. The demineralizers 80, 81, 82 are flow coupled to line 45 which is used to supply fresh resin to the demineralizer system 702. Line 70 is also provided for the introduction of sluice water to the demineralizer system 702 in a counter flow fashion to be used to flush spent resin out of the demineralizer vessels. The spent resin exits the demineralizer system 702 via line 71. The process fluid can also be diverted around the demineralizer system 702 via line 48. After the process fluids exit the demineralizer system 702 they are transported via line 44 to the resin fines filter system 703. The resin fines filter system 703 is provided to ensure that any resin from the demineralizer system 702 does not enter the primary system. The resin fines filter system 703 preferably contains a plurality of filters which have a combined total flow rate capacity equal to the decontamination system flow of about 1,500 gpm. In the preferred embodiment, four resin fines filters 306 are utilized. Each resin fines filter 306 has a capacity of 375 gpm and a filter rating of about 25 micron. After exiting the resin fines filter system 703 the process fluid is transported via line 47, and referring to FIG. 3, back to the primary system via line 58. Prior to the entry back into the primary system, chemicals are injected into the processed fluids. In the preferred embodiment, two chemical injection systems are utilized. First is a vanadous formate system 704. This vanadous formate system 704, in the preferred embodiment, has a vanadous formate tank 131 which contains the vanadous formate compounds in solution. The vanadous formate is preferably prepared by use of a recirculation and heater system shown as a vanadous formate mixing pump 206 flow coupled to a vanadous formate heater 91. When the vanadous formate solution is ready to be injected to the process fluids, the vanadous formate injection pumps 207, 208 are activated. This vanadous formate system 704 is utilized when a LOMI decontamination process is required by the decontamination process. The second chemical injection system is the chemical system 708 which is shown in the preferred embodiment as comprising a chemical mixing tank 132 which also has a recirculation and heater system shown as chemical mixing pump 209 and chemical heater 92 for dissolving decontamination chemicals in solution. When chemicals from the chemical mixing tank 132 are ready to be sent to the process, the chemical injection pumps 210 and 211 are activated. This chemical system 708 is designed to handle those chemicals utilized in a CAN-DEREM process as required by the decontamination process. In the preferred embodiment, both the vanadous formate 131 and the chemical mixing tank 132 are about 3000 gallons in size. The amount of decontamination chemicals which must be injected is dependent upon the amount of deposits in the RCS. The vanadous formate mixing pump 206 and the chemical mixing pump 209 are both preferably sized for a flow rate of about 100 gallons per minute. The vanadous formate injection pumps 207, 208 and the chemical injection pumps 210, 211 are preferably sized for a flow rate of 50 gallons per minute. The chemical injection system is flow coupled to the line 47 via line 49 for injection of the chemicals into the processed (demineralized and filtered) fluids prior to reentry of those fluids into the primary system via line 58. Referring now to FIG. 4, the new resin system 707 the spent resin storage system 705, the sluice water system 706, and the decontamination waste system 709 are shown. When the ion exchange resin is spent, the demineralizer system 702 has to be regenerated with new resin. The sluice water system 706 is employed to remove the spent resin. Sluice water is provided from the sluice water supply tank 113 to the demineralizer system via line 70. The sluice water travels through the sluice water pump 204 and the sluice water filter 310 prior to entering the demineralizer system 702. The sluice water system 706 also contains a sluice water recycle pump 205 for recycling the sluice water from the demineralizer system 702. The amount of sluice water required for transport of the resin is dependent upon the amount of resin. The sluice water supply tank is preferably sized to have a capacity of 1,800 gallons of sluice water. The sluice water filter 310 is preferably sized to have a capacity of 100 gpm and the sluice water pump 204 and the sluice recycle pump 205 are preferably designed to have a capacity of about 100 gpm flow rates. After the sluice water enters the demineralizer system 702 via line 70, the sluice water carries the spent resin from the demineralizer system 702 via line 71 to the spent resin storage system 705. The spent resin storage system 705 is comprised of a series of tanks which preferably have a combined total storage volume of about 34,400 gallons. The spent resin storage tanks 121 are provided with screen bottoms such that the sluice water exits these tanks and is recirculated via the sluice water recirculation pump 205 to the sluice water supply tank 113. Line 42, from the filtrate collection tank 101 is connected to line 71 upstream of the spent resin storage tank system 705. A new resin system 707 is preferably included in a decontamination process overall system in order to batch fresh resin to the demineralizer system 702. The new resin system 707 is preferably comprised of a resin supply tank 112 which contains fresh resin. This tank is flow coupled to line 66 which carries demineralizer water from demineralizer water source 65. The solution of resin and demineralizer water is sent to the resin batching tank 111 by the resin supply pump 203. A quantity of resin to fill a demineralizer vessel 611 is then transported from the resin batching tank 111 by the resin feed pump 202 via line 45. In the preferred embodiment, the resin supply tank 112 is capable of storing about 7000 gallons of resin. The resin batching tank 111 is preferably sized to hold about 2100 gallons of solution. The resin feed pump 202 and the resin supply pump 203 are both preferably sized to have a capacity of about 100 gpm. A decontamination waste system 709 is also provided. This decontamination waste system 709 comprises a decontamination waste tank 133 which preferably has a volume capacity of about 3000 gallons. The decontamination waste tank 133 is flow coupled to line 70. A decontamination waste pump 212 is flow coupled to the decontamination waste tank 133 for pumping the decontamination waste via line 73 to a storage system. The decontamination system 709 is designed to collect waste solutions from any of the group consisting of the backflush filter system 701, the demineralizer system 702, the spent resin storage tank 705, and the sluice water system 706. After the full system decontamination process was developed, the task of employing such a system outside of the containment chamber had to be met. Various design problems exist in implementing such a design, the most important being: (1) ensuring that the decontamination building can contain a spill if radioactive materials leak from the decontamination equipment; (2) locating the equipment so that it is properly shielded to protect personnel; (3) providing for ease of installation and disassembly of the decontamination equipment into and out of the decontamination building; (4) providing necessary safety equipment around the piping connecting the decontamination building to the RHR system; and (5) providing a modular design which can be easily stored in radioactive containers. The design layout configuration of the present invention provides for supplying the equipment necessary for the full system decontamination process in divisible units. These units are placed upon skids which are easily transported by tractor trailers to the reactor site. These units are also easily installed and dismantled and taken out of the decontamination building when not in use. Another key design feature was to design each system so that it could fit upon an individual skid, or a plurality of skids which would be situated in close proximity to one another. A further key design feature was the placement of the separate decontamination systems in separate high activity radioactive source shielded rooms. These rooms have an opening passageway of about from 1.5 ft. (0.5 m) to 3 ft. (0.9 m) to allow for personnel access, but which does not significantly affect the shielding protection afforded to the personnel. The chemical decontamination process on a full scale basis may only be needed two to three times per reactor life. The particular nuclear facility may therefore desire to remove the equipment when not in use. Therefore, a modular design was required which would lend itself to easy equipment set-up and removal. Referring now to FIG. 5, a preferred outside of containment layout design is shown for a typical four-loop pressurized water nuclear reactor. In this layout design, skid positions were established for various systems comprising the decontamination system. The back-flush filter system 701 is positioned on back-flush filter skid 501. The entire back-flush filter system 701 is not shown, however that entire system is preferably placed upon the back-flush filter system skid 501. The back-flush filter system 701 could optionally be placed on a multitude of skids. The entire back-flush filter system 701 is located inside a back-flush filter room 551 defined by the shield 93 and the building wall 94. The exact location of the back-flush filter room 551 is not critical, however the shielding around the back-flush filter system 701 is required. A small access way 95 is provided in order for personnel to maintain the back-flush filter system 701. The width of this access way 95 is about 1.5 ft. (0.5 m) to 3 ft. (0.9 m). In a preferred embodiment, all of the access ways 95 are of a labyrinth design such that no direct line of sight exists between the process equipment and the personnel. This design permits maximum personnel radiation protection and accessibility for maintenance. The demineralizer system 702 is shown as a multitude of demineralizer vessels 611 and any combination of such vessels is possible if the total volume capacity is sufficient. In the preferred embodiment each vessel is situated on an individual demineralizer skid 506. This type of layout is preferred for ease in set-up and removal of the equipment. The demineralizer system 702 is also contained within a high radioactivity source shielded room, preferably a separate room, the demineralizer system room 556. This demineralizer system room 556 is also defined by the shield 93. The exact location of the demineralizer system room 556 is not critical, however the shielding of the entire demineralizer system 702 is required. The access way 95 allows for personnel to maintain the demineralizer system 702. It is possible to house each individual demineralizer skid 506 in a separate shielded room, or combine a few vessels, say three vessels, in a separate shielded room. Such a multi-room design for the demineralizer system 702 would allow for the separation of a single, or a battery, of vessels containing high activity radioactive materials or for the separation of leaking vessels while the other vessels could still be operated. In the preferred embodiment, each demineralizer vessel 611 is placed on an individual skid to allow for storing the vessel in a safety container upon removal of the decontamination system when the system is not in use. The demineralizer skid 506 is designed to fit into a low specific activity (LSA) container for transport and storage. These containers are obtained from manufacturers and currently have a maximum width of 8 ft. (2.4 m) and height of 9 ft. (2.7 m). The maximum length, height, and width is dictated by designing the system to be easily transported on the roadways without requiring a special transportation permit. Therefore, the skids are sized to be about 7.5 ft. (2.3 m) in width and about 9 ft. (2.7 m) in length with the height defined by the vessel. In transporting the vessels in the LSA containment, the skid is placed on its side, if necessary, to fit into the LSA container. The height is preferably kept below 40 ft (12.2 m) to ensure that no special transportation permit is required for these vessels. The resin fines filter system 703 is positioned on the resin fines filter skid 502. The entire resin fines filter system 703 is not shown, however that entire system is preferably placed upon the resin fines filter system skid 502. The resin fines filter system 703 could optionally be placed on a multitude of skids. The entire resin fines filter system 703 is preferably located inside a resin fines filter room 552 defined by the shield 93 and the building wall 94. The exact location of the resin fines filter room 552 is not critical, however the shielding around the resin fines filter system 703 is required. A small access way 95 is provided in order for personnel to maintain the resin fines filter system 703. The spent resin storage system 705 is shown as a multitude of spent resin storage tanks 121 and any combination of such tanks is possible if the total volume capacity is sufficient. In the preferred embodiment each tank is situated on an individual spent resin storage system skid 509. This type of layout is preferred for ease in set-up and removal of the equipment. The spent resin storage system 705 is also preferably contained within a separate high radioactivity source shielded room, the spent resin storage room 559. This spent resin storage room 559 is defined by the shield 93 and the building wall 94. The exact location of the spent resin storage room 559 is not critical, however the shielding of the entire spent resin storage system 705 is required. The access way 95 allows for personnel to maintain the spent resin storage system 705. It is possible to house each individual spent resin storage system skid 509 in a separate shielded room, or combine a few tanks, say two tanks, in a separate shielded room. Such a multi-room design for the spent resin storage system 705 would allow for the separation of a single, or a battery, of tanks containing high levels of radioactive materials or for the separation of leaking tanks while the other tanks could still be operated. In the preferred embodiment, the chemical decontamination equipment is removed when not in use. The spent resin storage tanks 121, as with the demineralizer vessels 611, are therefore stored in LSA containers, which prevent personnel exposure to any radioactive material on the spent resin storage tanks 121. The placement of each spent resin storage tank 121 on a separate spent resin storage system skid 509 is preferred to facilitate this storage process. The spent resin storage system skid 509 was designed with the same skid dimensions of the demineralizer skids 506. The sluice water system 706 is positioned on the sluice water system skid 504. The entire sluice water system 706 is not shown, however that entire system is preferably placed upon the sluice water system skid 504. The sluice water system 706 could optionally be placed on a multitude of skids. The entire sluice water system 706 is preferably located inside a sluice water system room 554 defined by the shield 93 and the building wall 94. The exact location of the sluice water system room 554 is not critical, however the shielding around the sluice water system 706 is required. A small access way 95 is provided in order for personnel to maintain the sluice water system 706. The decontamination waste system 709 is positioned on the decontamination waste system skid 503. The entire decontamination waste system 709 is not shown, however that entire system is preferably placed upon the decontamination waste system skid 503. The decontamination waste system 709 could optionally be placed on a multitude of skids. The entire decontamination waste system 709 is preferably located inside a decontamination waste system room 553 defined by the shield 93 and the building wall 94. The exact location of the decontamination waste system room 553 is not critical, however the shielding around the decontamination waste system 709 is required. A small access way 95 is provided in order for personnel to maintain the decontamination waste system 709. The chemical injection systems are preferably placed within close proximity to the other chemical decontamination systems which handle the radioactive material. In FIG. 6, the vanadous formate system 704 is shown as being located on a vanadous formate skid 513. In the most preferred embodiment, the vanadous formate system 704 is located on one skid, however a multitude of such skids could be utilized. This vanadous formate system 704 does not have to be shielded because no radioactive materials are processed by this system. The chemical system 708 is shown as being located on a chemical skid 514. In the most preferred embodiment, the chemical system 708 is located on one skid, however a multitude of such skids could be utilized. This chemical system 708 does not have to be shielded because no radioactive materials are processed by this system. The new resin system 707 is shown as being located on a new resin system skid 505. In the most preferred embodiment, the new resin system 707 is located on one skid, however a multitude of such skids could be utilized. The new resin system 707 does not have to be shielded because no radioactive materials are processed by this system. The skid dimensions for the back-flush filter system 702, the resin fines filter system 703, the sluice water system 706, and the decontamination waste system 709 are preferably designed to be stored in LSA containers and be transportable without a special road transportation permit. Therefore, the skids are designed to be less than 8 ft. (2.4 m) wide. The height of the equipment on these skids is kept below 9 ft. (2.7 m). The length of the skids is kept below about 40 ft. (12.2 m) in order for the entire skid/LSA container to be transported over the roadway without a special transportation permit. Various types of shielding walls can be used for the shield 93 and the building wall 94. Typical shielding walls vary from 8 inches (20.3 cm) to 32 inches (81.3 cm) thick solid block walls made of concrete. These walls can also contain a lead additive intermixed with the concrete. These shielding walls are capable of protecting personnel from a high radioactive source such as the equipment which processes the primary process fluids during the chemical decontamination process. |
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059360070 | claims | 1. A medical article of manufacture prepared from a polycarbonate moulding composition comprising: a) 97.5 wt. % to 99.9 wt. % of a polycarbonate or a copolycarbonate; b) 0.1 wt. % to 2.5 wt. % of a .gamma.-radiation stabiliser of the formula (I), each percentage by weight being with reference to 100 wt. % of a)+b), ##STR8## where R.sub.1 and R.sub.2, independently of each other, represent a methyl group, a C.sub.7 -C.sub.18 optionally branched and/or substituted alkylaryl group or a C.sub.6 aryl group, R.sub.4 represents a methyl group, a C.sub.7 -C.sub.18 optionally branched and/or substituted alkylaryl group, a C.sub.6 aryl group or H, a) 97.5 wt. % to 99.9 wt. % of a polycarbonate or a copolycarbonate; and b) 0.1 wt. % to 2.5 wt. % of a .gamma.-radiation stabiliser of the formula (I), each percentage by weight being with reference to 100 wt. % of a)+b), ##STR11## where R.sub.1 and R.sub.2, independently of each other, represent a methyl group, a C.sub.7 -C.sub.18 optionally branched and/or substituted alkylaryl group or a C.sub.6 aryl group, R.sub.4 represents a methyl group, a C.sub.7 -C.sub.18 optionally branched and/or substituted alkylaryl group, a C.sub.6 aryl group, or H, 2. The medical article of manufacture according to claim 1, wherein the polycarbonate is prepared from 2,2-bis-(4-hydrozypenyl)-propane, 1, 1-bis-(4-hydroxyphenyl)-3,3,5-trimethylcyclohexane, 1,1-bis(4-hydroxyphenyl)-1-phenylethane or a mixture thereof. 3. The medical article of manufacture according to claim 1, wherein the .gamma.-radiation stabiliser corresponds to the formula (Ia): ##STR10## where R.sub.1 and R.sub.2, independently of each other, represent a methyl, a phenyl or a cresyl group, 4. The medical article of manufacture according to claim 1, wherein the polypropylene glycol of step c) has an average molecular weight of 800 to 4,000. 5. The medical article of manufacture according to claim 1, wherein the polypropylene glycol of step c) is present in an amount of 0.1 wt. % to 1.5 wt. %, the percentages by weight of the polypropylene glycol each being with reference to 100 wt. % of a)+b). 6. The medical article of manufacture according to claim 1, the polycarbonate moulding composition further comprising a phosphorous containing stabiliser. 7. The medical article of manufacture according to claim 1, the polycarbonate moulding composition further comprising a thermoplastic material. 8. The medical article of manufacture according to claim 7, wherein the thermoplastic material is present in a amount of 10 wt. % to 50 wt. %, the percentages by weight of the thermoplastic material each being with reference to 100 wt. % of a)+b). 9. The medical article of manufacture according to claim 7, wherein the thermoplastic material comprises an aromatic polyester carbonate, a polyalkylene terephthalate, a EPDM polymer, a polystyrene, a copolymer based on styrene, and a polycarbonate based on different bisphenol from the polycarbonate according to claim 7. 10. The medical article of manufacture according to claim 1, the polycarbonate moulding composition further comprising an additive. 11. The medical article of manufacture according to claim 10, wherein the additive comprises a thermal stabiliser, an UV stabiliser, an optical brightener, a flame retardant, a mould release agent, a colorant, a pigment, an antistatic agent, a filler, a reinforcing substance or a mixture thereof. 12. The medical article of manufacture according to claim 1, which comprises a dialyser housing. 13. A dialyser housing prepared from a polycarbonate moulding composition comprising: |
claims | 1. Process for manufacturing a nuclear component via the method of chemical vapor deposition of an organometallic compound by direct liquid injection (DLI-MOCVD), the nuclear component chosen from a nuclear fuel cladding, a spacer grid, a guide tube, a plate fuel or an absorber rod, comprising:i) a support containing a substrate based on a metal chosen from zirconium, titanium, vanadium, molybdenum or base alloys thereof (1) and at least one protective layer (2);ii) said at least one protective layer (2) coating said support and composed of a protective material comprising chromium which is an amorphous chromium carbide; the process comprising the following successive steps:a) vaporizing a mother solution containing a hydrocarbon-based solvent free of oxygen atoms; and a precursor of bis(arene) type comprising chromium, the precursor having a decomposition temperature comprised between 300° C. and 600° C.;b) in a chemical vapor deposition reactor in which is located said support to be covered and the atmosphere of which is at a deposition temperature comprised between 300° C. and 500° C. and at a deposition pressure comprised between 13 Pa and 7000 Pa; introducing the mother solution vaporized in step a), which brings about the deposition of said at least one protective layer (2) on said support. 2. Process for manufacturing a nuclear component according to claim 1, wherein the nuclear component further comprises a liner (4) placed on the inner surface of said support, which is the surface of said support opposite to the medium that is external to the nuclear component. 3. Process for manufacturing a nuclear component according to claim 1, wherein said at least one protective layer (2) is an outer protective layer (2A) which coats the outer surface of said support which is the surface of said support facing the medium that is external to the nuclear component; and/or, when the nuclear component comprises an inner volume, an inner protective layer (2B) which coats the inner surface of said support. 4. Process for manufacturing a nuclear component according to claim 2, wherein the liner (4) is deposited, at a deposition temperature comprised between 200° C. and 400° C., onto the inner surface of said support by chemical vapor deposition of an organometallic compound (MOCVD) or DLI-MOCVD with, as precursor(s), a titanium amide and further a precursor comprising silicon, a precursor comprising aluminum and/or a liquid additive comprising nitrogen if the material of which the liner is composed comprises, respectively, silicon, aluminum and/or nitrogen. 5. Process for manufacturing a nuclear component according to claim 1, wherein the process comprises, after step b):c) performing on said at least one protective layer (2) at least one step chosen from a subsequent treatment step of ionic or gaseous nitridation, ionic or gaseous silicidation, ionic or gaseous carbosilicidation, or ionic or gaseous nitridation followed by ionic or gaseous silicidation or carbosilicidation. 6. Process for manufacturing a nuclear component according to claim 1, wherein the mother solution further contains an additional precursor having a decomposition temperature comprised between 300° C. and 600° C., the additional precursor being at least one precursor of bis(arene) type comprising an addition element chosen from yttrium, aluminum, vanadium, niobium, molybdenum, tungsten, a precursor comprising aluminum or yttrium as addition elements, or mixtures thereof; such that the protective material is doped with the addition element. 7. Process for manufacturing a nuclear component according to claim 1, wherein the precursor of bis(arene) type comprising chromium further comprises an element M which is, chromium; the element M being in oxidation state zero (M0) so as to have a precursor of bis(arene) type comprising the element M0. 8. Process for manufacturing a nuclear component according to claim 1, wherein the substrate (1) is coated with an interposed layer (3) placed between the substrate (1) and the at least one protective layer (2). 9. Process for manufacturing a nuclear component according to claim 2, wherein said at least one protective layer (2) is an outer protective layer (2A) which coats the outer surface of said support which is the surface of said support facing the medium that is external to the nuclear component; and/or, when the nuclear component comprises an inner volume, an inner protective layer (2B) which coats the inner surface of said support coated with the liner (4). 10. Process for manufacturing a nuclear component according to claim 6, wherein the precursor of bis(arene) type comprising the addition element comprises an element M which is the addition element; the element M being in oxidation state zero (M0) so as to have a precursor of bis(arene) type comprising the element M0. |
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abstract | Improved image reconstruction methods for cone beam ROI (region of interest) imaging of long objects with an area detector using a two or more circular scans and a connecting line scan, wherein essentially the entire detector area is utilized for acquiring cone beam data for all source positions. |
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abstract | Example embodiments are directed to core spray sparger T-box repairs, specifically, to universal core spray sparger T-box weldless clamps having remote-friendly operation and methods of using universal core spray sparger T-box weldless clamps. Example embodiment clamps may be secured without welding to a variety of upper and lower sparger T-box configurations. Example embodiment clamps may be configured to simultaneously engage a sparger T-box in multiple dimensions to allow a universal fit. Further, example embodiment clamps may be accessed, installed, or removed by interacting only with a front side of the example embodiment clamps, thus potentially reducing difficulty and cost in remote access repairs to example clamps. |
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abstract | The present disclosure provides an X-ray imaging apparatus and control method thereof, for guiding the user to intuitively recognize an actual dose of X-rays and select a proper dose, ultimately a condition for low dose of X-ray irradiation by providing the user with information about an actual X-ray dose to which an X-ray filter effect is reflected. In accordance with an aspect of the disclosure, an X-ray imaging apparatus includes: an X-ray source configured to generate and irradiate X-rays according to an X-ray irradiation condition including at least one of a tube voltage, a tube current, or a filter; a display configured to provide a graphic user interface to receive a choice about the X-ray irradiation condition; and a controller configured to obtain a parameter that represents a dose of radiation, to which an influence of the filter is reflected, based on the selected X-ray irradiation condition and control the display to display the parameter. |
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description | As shown in FIG. 1, a nuclear power station conventionally includes a reactor pressure vessel 10 sealed within a containment structure 50 that houses several power-producing systems and equipment. Reactor 10 may include various configurations of fuel and reactor internals for producing nuclear power. For example, vessel 10 may include several fuel assemblies positioned within a general cylindrical core. Fluid coolant and/or moderator may flow through reactor 10; for example, in US light water reactors, the fluid may be purified water, in natural uranium reactors, the fluid may be purified heavy water, and in gas-cooled reactors, the fluid coolant may be a gas such as helium, with moderation provided by other structures. Vessel 10 may be sealed and opened through upper head 95 at flange 90. As shown in FIG. 1, during plant fabrication and at regular service and/or refueling outages, upper head 95 may be removed and operators and/or equipment can access internals of vessel 10 inside of containment structure 50 for various purposes. For example, with access to the reactor internals, some of fuel bundle assemblies may be replaced and/or moved between within the core and a fuel staging or spent fuel pool area(s), and maintenance/installation on other reactor structures in containment 50 may be performed. During such maintenance, a refueling cavity 20 above flange 90 and surrounding reactor 10 may be filled, or flooded, with fluid coolant. The fluid coolant may both remove heat and block radiation from escaping to operators around cavity 20, such as workers performing maintenance on operations floor 25 above cavity 20. With such shielding, refueling cavity 20 may be used for storage of radioactive structures and a staging area for fuel handling, as well as a general interface for access into reactor 10. Refueling bridge 1 with mast 3 and grapple 4 are useable during outages with access to reactor vessel 10 to perform fuel offloading, reloading, shuffling, and/or maintenance. Refueling bridge 1 may be positioned on operations floor 25 above or about flange 90 when reactor vessel 10 is opened. Bridge 1 may include a trolley 2 capable of rotating and/or laterally moving to any horizontal or vertical position. Trolley 2 may include a refueling mast 3 with hoist box and grapple 4 that descend into reactor 10 and perform fuel and other structure movements throughout cavity 20 during outages. At other outage periods and during operations, cavity 20 may be drained completely or partially (such as down to flange 90). Because cavity 20 may have previously been flooded with fluid coolant before such draining, residues from and particulates in the fluid coolant may adhere to cavity surfaces, including cavity walls 21. These remnants from the fluid may be undesirable—such as radioactive or chemically corrosive—for operating conditions within cavity 20, on operations floor 25, and/or anywhere throughout containment building 50. As such, operators sometimes take measures to reduce particulates and impurities in any fluid that fills cavity 20. For example, plant operators may add solvents or otherwise change coolant chemistry to reduce deposition on surfaces drained of coolant and/or may use submersible, stationary filters on a floor of cavity 20. For example, underwater filters from Tri Nuclear Corporation may sit on a bottom of cavity 20 and filter or demineralize fluid in cavity 20. Example embodiments and methods reduce settling of unwanted materials out of a fluid onto structures by causing a flow around the structures. Example embodiments use a fluid source and discharge the fluid from the source against the structure in the fluid. The discharged fluid flow and ambient fluid surrounding the structure may be the same or different. For example, both fluids may be water, but the ambient fluid may have unwanted particulate or dissolved contaminates in it, whereas the sprayed fluid might be filtered and/or chemically treated to help remove the unwanted materials from the structures. Example embodiments and methods may use a flow rate of approximately 2 meters per second or more, which is effective in several types of water to prevent deposition out of the water onto surfaces. The flow rate may be created by a pressurized fluid source and/or a local pump, and the jetted fluid may come from the same volume surrounding the structure, but with optional filtering, temperature adjustment, and/or chemical treatment, for all or a portion of the fluid jet. Example embodiments may be wholly submerged in the fluid and still operate, using the flow discharge to move in the fluid to spray different desired surfaces, as well as other movement methods like changing buoyancy. Example embodiment systems may also work with portions in the fluid and other portions outside the fluid. For example, a multi-stage filter may be fitted inside a mobile assembly and submerged in coolant water in a flooded cavity, where the water is passed through the filter and dispersed to create the 2 m/s rate by an induction pump. Alternatively or additionally, another filter and pump may suck the water coolant from the cavity and feed it through a base outside the cavity where the water is treated chemically and thermally and delivered back into the cavity to be sprayed at deposition surfaces. With proper buoyancy, sizing, and spray discharge, any submerged mobile assembly may move between or to desired surfaces to be cleaned. Multi-stage filters useable with example embodiments may remove a variety of contaminants, including metallic conjugates specifically liberated by the water chemistry of the flow. Example embodiment filters may include coarse reservoirs, fibrous filters, charged particles, sintered metallics, resins, etc. in several different stages that are independently removable and disposable. This is a patent document, and general broad rules of construction should be applied when reading and understanding it. Everything described and shown in this document is an example of subject matter falling within the scope of the appended claims. Any specific structural and functional details disclosed herein are merely for purposes of describing how to make and use example embodiments or methods. Several different embodiments not specifically disclosed herein fall within the claim scope; as such, the claims may be embodied in many alternate forms and should not be construed as limited to only example embodiments set forth herein. It will be understood that, although the terms first, second, etc. may be used herein to describe various elements, these elements should not be limited by these terms. These terms are only used to distinguish one element from another. For example, a first element could be termed a second element, and, similarly, a second element could be termed a first element, without departing from the scope of example embodiments. As used herein, the term “and/or” includes any and all combinations of one or more of the associated listed items. It will be understood that when an element is referred to as being “connected,” “coupled,” “mated,” “attached,” or “fixed” to another element, it can be directly connected or coupled to the other element or intervening elements may be present. In contrast, when an element is referred to as being “directly connected” or “directly coupled” to another element, there are no intervening elements present. Other words used to describe the relationship between elements should be interpreted in a like fashion (e.g., “between” versus “directly between”, “adjacent” versus “directly adjacent”, etc.). Similarly, a term such as “communicatively connected” includes all variations of information exchange routes between two devices, including intermediary devices, networks, etc., connected wirelessly or not. As used herein, the singular forms “a”, “an” and “the” are intended to include both the singular and plural forms, unless the language explicitly indicates otherwise with words like “only,” “single,” and/or “one.” It will be further understood that the terms “comprises”, “comprising,”, “includes” and/or “including”, when used herein, specify the presence of stated features, steps, operations, elements, ideas, and/or components, but do not themselves preclude the presence or addition of one or more other features, steps, operations, elements, components, ideas, and/or groups thereof. It should also be noted that the structures and operations discussed below may occur out of the order described and/or noted in the figures. For example, two operations and/or figures shown in succession may in fact be executed concurrently or may sometimes be executed in the reverse order, depending upon the functionality/acts involved. Similarly, individual operations within example methods described below may be executed repetitively, individually or sequentially, so as to provide looping or other series of operations aside from the single operations described below. It should be presumed that any embodiment having features and functionality described below, in any workable combination, falls within the scope of example embodiments. The inventors have recognized that existing coolant cleanup in nuclear power facilities, focusing on ion removal from reactor coolant with stationary scrubbers in a flooded cavity and/or through existing coolant clean-up filters, resins, and deionizers in combination with adjusting coolant chemical properties to decrease particulate deposition, does not fully remove complexed metal ions present as particulates in reactor coolant. This is especially problematic with metallic radioisotopes like Cobalt, Cesium (particularly in the case of a fuel rod leak), and Zinc, which readily complex with Iron to form particulates that deposit on flooded surfaces and cannot be effectively removed with conventional mechanical and chemical remediation measures. These radioisotopes deposited on flooded surfaces generally remain over time and can become airborne upon dry-out, presenting a significant radiation dose contribution to personnel and equipment in the areas during and after flooding, as well as serving as a reservoir for future coolant contamination when disturbed from the surfaces by re-flooding. The inventors have recognized that radioactive, complexed metallic particulates that have deposited on plant surfaces during contact with coolant may be removed through fluid-mechanical action. On deposition surfaces, particulates are generally not exposed to higher fluid flows because of the nature of the boundary layer adjacent to a stationary surface formed in a reactor coolant like water. However, by causing fluid flows of sufficient velocity, the metallic particulates can be removed from the surface and prevented from re-depositing on the surface. Thus, by moving coolant at a sufficient transport velocity at deposition surfaces, metallic particulates may be kept in the coolant where they can be removed through conventional scrubbing and/or additional filtering, preventing them from depositing and increasing radiation exposure. In order to discover the necessary transport velocity to avoid the newly-recognized fluid-dynamic solution to radioisotope deposition, the inventors looked to Poirier, “Minimum Velocity Required to Transport Solid Particles from the 2H-Evaporator to the Tank Farm” US DoE Technical Report WSRC-TR-2000-00263, Sep. 27, 2000, incorporated by reference herein in its entirety, as a reference for particulate transport velocities in closed systems. Repurposing the transport and settling velocity calculations from the Poirier report for open systems with the density characteristics of Cobalt particulates and using typical diameters of such particulates to derive Reynolds numbers in the solutions, the inventors discovered that a flow rate of about 2 meters per second inhibited deposition of particulates up to 5 millimeters in diameter. This rate is well below the expected necessary rate for particulate transport speed, especially in light of its use in an open system and compared to the velocities in the Poirier report. The inventors further recognized that movement of fluid at speeds well below 2 m/s at deposition surfaces results in high levels of settling of radioactive particulates. The below disclosure uniquely overcomes these and other problems, by leveraging systems and methods that move particulate-bearing fluid at calculated speeds near or above 2 m/s, sufficient to prevent settling of radioactive contaminants on these surfaces. The present invention is systems and methods of reducing and/or preventing unwanted depositions on surfaces by creating fluid flows on those surfaces above a settling velocity of the unwanted substances. In contrast to the present invention, the few example embodiments and example methods discussed below illustrate just a subset of the variety of different configurations that can be used as and/or in connection with the present invention. FIG. 2 is an illustration of an example embodiment flow inducer system 100 usable to prevent particulate settling on surfaces in fluid coolant, including preventive radioactive particle deposition on structures immersed in fluids bearing the same. As shown in FIG. 2, system 100 may include a base 120 positioned about or above a reactor 10. For example, inducer system 100 may be positioned about flange 90 of reactor 10 during a maintenance outage in which an upper head of reactor 10 is removed for access to the fuel core and reactor internals. Example embodiment system 100 includes a mobile assembly 150 that can extend down into reactor 10 and into coolant therein, via a connection 101. In this way, example embodiment flow inducer system 100 may include components and/or operations interfaces in base 120, such as electrical power connections, user interfaces, purified coolant sources, external movement structures, etc., that function best outside of coolant, while mobile assembly 150 induces flow in coolant in which it is immersed and remote from base 120. Alternatively, it is understood that base 120 may be combined into mobile assembly 150 to provide a unitary structure for inducing flow and preventing particulate deposition on surfaces exposed to coolant or other particulate contaminant-bearing fluids. Mobile assembly 150 causes coolant flow of approximately 2 meters per second or more to be directed to desired surfaces. Mobile assembly 150 is moveable within the coolant along surfaces and in spaces containing the same to prevent deposition at several positions. For example, as shown in FIG. 2, mobile assembly 150 may move vertically along connection 101 to reach several different axial positions of a wall of reactor 10. Similarly, mobile assembly 150 may move radially or angularly with proper forces to any other surface at which an induced flow may be desired to reduce deposition. A track 190 or other movement path, such as one provided via crane or other locomotive structures, may be provided about flange 90 to permit angular movement of base 120 as well. Similarly, track 190 could be positioned on an operations floor 25 or other area to provide desired movement and/or positioning of example embodiment flow inducer system 100. Although example embodiment flow inducer system 100 is shown in FIG. 2 about a reactor 10 at flange 90, it is understood that inducer 100 may be installed at other locations. For example, a base 120 could be positioned on a containment operations floor 25 (FIG. 1), with mobile assembly 150 extending into and moving within a cavity 20 (FIG. 1). Or, for example, system 100 may be used in a spent fuel pool or new fuel staging area within a nuclear power plant. Still further, example embodiment flow inducer system 100 may be used in any system with fluid contamination removable through fluid flow. FIGS. 3 and 4 are illustrations of portions of example embodiment flow inducer systems, with FIG. 3 illustrating in detail an example embodiment mobile assembly and FIG. 4 illustrating in detail an example embodiment base. As such, the components of FIG. 3 could be useable as or with mobile assembly 150 of FIG. 2, and components of FIG. 4 could be useable as or with base 120 of FIG. 2. Or example embodiments of FIGS. 3 and 4 may be used separately or with different systems or combined into a single mobile system, for example. As shown in FIG. 3, an example embodiment mobile assembly 300 may include a variety of components to create a deposition-reducing coolant flow against several different surfaces in a volume of coolant fluid. Assembly 300 may be connected to a guide or movement arm for positioning. For example, a wire, pole, or removably fixed rod 301 may span an axial depth of a refueling cavity 20 or other space, and example embodiment assembly 300 may connect to rod 301 through a movable connector 302 like a keyhole, loop, or grommet that permits only axial movement of assembly 300 along fixed rod 301. Rod 301 may connect to other components outside of cavity 20, such as a base 120 including coolant supply 354, electrical supply 358, etc., or rod 301 may be used in isolation. Coolant supply 354 may be coupled with pole 301 or otherwise supplied to mobile assembly 300. Coolant supply 354 may provide additional volume of coolant or other compatible fluid for creating induced flow for removing particulates. For example, if coolant is deionized or borated light water, coolant supply 354 may supply matching water. Coolant supply 354 may also provide relatively cleaner fluid as well as chemically-treated and temperature-moderated fluid for optimal contaminate clean-up. For example, coolant supply 354 may provide relatively colder water treated with a weak acid and/or oxidizer to enhance particulate solubility and removability by filters. Coolant supply 354 may also provide coolant for inducing fluid flow at a higher or operating pressure for example embodiment mobile assembly 300. Coolant supply 354 may feed directly into assembly 300 or connect via a coolant supply connection 355, which may be tubing or an injector, for example. Example embodiment mobile assembly 300 may also include a pump 344 or other hydrodynamic flow-inducing structure. For example, pump 344 may be an inductive jet pump, a centrifugal pump, a hydraulic pump, etc. Pump 344 may be locally powered through batteries or may be connected to an external, remote power source, such as electrical supply 358 via rod 301. Although pump 344 may be omitted with sufficient pressure and flow shaping from coolant supply 354 to create desired coolant flows, pump 344 may be used with a pressurized coolant supply 354 or without coolant supply 354. Example embodiment mobile assembly 300 may use fluid provided from coolant supply 354 and/or coolant from cavity 20 to create a flow directed at desired surfaces, such as cavity wall 21. In the example of FIG. 3, assembly 300 uses both provided and ambient coolant fluid in creating a flow 352. For example, lower-temperature coolant from coolant supply 354 may enter an upper manifold 356 and flow down through a series of tubes and/or baffles in a heat exchanger 357. The coolant may flow into a lower collection manifold 358 from the tubes and into a final section of coolant supply connection 355, which may be a flexible tube or injection device. Pump 344 then pressurizes and accelerates the coolant, potentially through a nozzle and/or diffuser, into an induced flow 352 against surfaces 21. Additionally, ambient coolant from cavity 20 may be taken in through a top inlet 353 and passed through an internal filter 350 around heat exchanger 357. Internal filter 350 may filter out impurities and dislodged/dissolved radionuclide depositions from ambient coolant taken from cavity 20, permitting relatively cleaner induced fluid flows. An example embodiment filter useable as filter 350 is discussed in connection with FIG. 5. If coolant from coolant supply 354 is colder than coolant in cavity 20, natural convection from the lower-temperature coolant in heat exchanger 357 may aid in driving ambient coolant from cavity 20 into inlet 353 and internal filter 350. Ambient coolant, after being filtered through internal filter 350, may connect to pump 344 through an ambient coolant connector 345. Pump 344 may entrain ambient coolant from ambient coolant connector 345 with any accelerated coolant provided from coolant supply 354 via coolant supply connection 355. With the use of a proper flow path, potentially including a diffuser, accelerated coolant from pump 344 may provide a suction to ambient coolant connector 345, drawing additional ambient coolant into inlet 353 and through filter 350. For example, with proper pump power and flow path, coolant may be drawn from ambient coolant at a 2-to-1 ratio of coolant from coolant supply 354. Although example embodiment mobile assembly 300 uses both provided coolant and ambient coolant to create a coolant flow with a pump, it is understood that other combinations are useable in example embodiments. For example, only a pressurized coolant source and nozzle may be used to generate a desired coolant flow without a pump or filter. Or, for example, only a locally-powered pump and ambient coolant may be used to create coolant flows without need for external sources. Or, as shown in FIG. 3, all systems may be used together. Induced coolant flow 352 is ejected or discharged under the force of pump 344 and potentially a nozzle or diffuser at any desired velocity. For example, with proper pump power and/or flow path narrowing, coolant flow 352 may be 2 m/s or greater, resulting in desired deposition preventing and removing discussed above. Coolant flow 352 may be directed at various surfaces desired to be keep free from radionuclide deposition while immersed in coolant, such as cavity wall 21. Example embodiment mobile assembly 300 may also be moveable, axially or otherwise, due to coolant flow 352. For example, if the coolant is light water in a flow 352 into a flooded cavity 20 of the same, sufficient force may be generated by flow 352 on assembly 300 to move assembly 300 upward along pole 301, even with flow 352 at only a slight downward angle. Flow 352 may be redirected and/or changed in intensity to create desired upward or downward movement of mobile assembly 300 along pole 301, potentially reaching an entire axial length of a surface positioned nearby under only the forces generated by coolant flow 352. Similarly, gravity and buoyancy may be used to selectively move example embodiment mobile assembly 300 in a sufficiently dense coolant like water, alone or in combination with forces from flow 352, as well as other movement structures and forces. Sufficient upward movement axially may also enhance ambient coolant flow into inlet 353 for filtering, if used. As shown in FIG. 4, an example embodiment base 400 may include a variety of components to treat and provide fluid coolant to, and potentially move and control, a mobile assembly for creating flow. Example embodiment base 400 may be positioned near or above a coolant-filled space to be jetted or exposed to deposition-removing flows by mobile assembly 150, such as refueling cavity 20 for example. Or base 400 may be more distantly located, potentially spread among several different facilities, or a component within mobile assembly 150. Coolant may be provided to base 400 from any source, including a flooded cavity 20, coolant reserve, plant feedwater, local taps, etc. For example, a suction filter 410 may be immersed in coolant in cavity 20, and coolant may be drawn into base piping 411 through filter 410 by a pump 413. Filter 410 may effectively remove radionuclides in solution or as particulates in coolant. For example, filter 410 may be similar to example embodiment filter 500 discussed in connection with FIG. 5 useable in an example embodiment mobile assembly 300. Piping 411 may be any transport path capable of carrying fluid coolant, including plastic tubing and metal pipes. Pump 413 may be any type of fluid-motive device, including those designs useable as pump 344 (FIG. 3) in an example embodiment mobile device as well as larger or non-submergible pumps that work outside of a coolant. Example embodiment base 400 may include several components for creating optimal coolant to supply to mobile assembly 150, including optimal cleanliness, optimal temperature, and/or optimal chemistry. For example, a heat exchanger 412 may be placed along piping 411 at any point to substantially reduce a temperature of coolant, such that coolant provided to mobile assembly 150 is lower than ambient coolant temperature and can be used for natural convective movement and/or reduce deposition potential with lower temperature. And, for example a chemical injector system 420 may be installed along piping 411 to provide desired pH, buffering, oxidation, oxygenation, boration, surfactant, clarity, salinity, replacement cations, and/or resin concentration, etc. to coolant. As shown in FIG. 4, an example of a chemical injector system 420 may include a venturi 421 installed along piping 411. The low-pressure pinch point of venturi 421 may provide a suction for chemicals to be injected into the coolant at that point when a stop valve 422 is opened. Similarly, an injector or flow mixer may be used for venturi 421 to provide desired additions to coolant. Beyond stop valve 422 may be several different additive tanks with their own stop valves to control specific types of additives. For example, a pre-oxidizer, such as hydrogen peroxide, may be held in tank 424 by valve 423, and a dilute acid, such as a relatively weaker nitric acid, may be held in tank 426 by valve 425. By mixing the components of tanks 424 and 426 in desired proportions and total amounts through valves 423, 425, and 422, water used as coolant may include a dilute acid that catalyzes or accelerates oxidation reactions within surfaces exposed to induced flows including the acid. Local water coolant pH in the range of 5-6 can be maintained near such surfaces to in this way, facilitating metallic deposition removal and dissolution. Metal-enriched oxides on the surfaces can further be oxidized by hydrogen peroxide in the water coolant to a soluble ion, such as oxidizing chromium-based oxides to soluble chromates, under these conditions. Radionuclides in the oxides may thus be more readily removed through filters in example embodiment systems as well as in existing coolant cleanup systems. Of course, other desired chemicals may be injected through any number of different tanks to achieve desired coolant flow chemistry. Example embodiment base 400 may connect to a mobile assembly 150 through connection 101, providing treated coolant at a desired pressure for use in creating a flow to prevent particulate settling. Similarly, electrical power, operator instructions, and/or relocation/locomotion may be provided through connection 101 from base 400. FIG. 5 is an illustration of an example embodiment filter 500 useable as filter 350 in example embodiment mobile assembly 300 (FIG. 3) and/or base filter 410 (FIG. 4). As shown in FIG. 5, filter 500 may include several different layers configured to filter out unwanted coolant impurities, including radionuclides in a metallic complexes dissolved in the coolant, potentially after being removed from a surface deposition in the coolant by example embodiment systems. The layers may be discreetly staged or progressive to filter finer and finer contaminants. For example, just below inlet 353 (FIG. 3), may be a coarse reservoir 534 with wide-pitch filters to stop macro objects like filings, paint chips, fasteners, rags, etc. that often fall into coolant spaces during maintenance. A fibrous filter 533 may be next with denser mesh or fibrous layers that catch large particulates in the coolant. Below may be a charged bed 532 of a material with an electrostatic potential, like a sand or fine gravel with varying surface ions or charged polymer chains, that attracts and holds smaller corrosion particles out of the coolant passing therethrough. A metallic filtering bed 531 may be placed next with sintered or finely-porous corrugated metallic sheets. Finally a resin bed 530 may be captured between two screens 529 and 528. Resin bed 530 may be a non-soluble ionized resin, like those used in conventional nuclear power coolant polishing and cleanup systems. These resins may include known products like Amberlyte, cross-linked polystyrenes, and Amberjet. Resin bed 530 may be specifically matched to capture known metallic complexes released into coolant following exposure of a contaminated surface to a flow rate of a transport velocity. Screens 529 and 528 may be sufficiently fine to prevent resin from migrating out of filter 500 while allowing clean coolant to freely pass. A backup screen 527 may be below screen 528 to prevent escape of resin 530 in the case of failure of screen 528. Coolant may flow through each filter stage 534, 533, 532, 531, and 530 progressively, into collector 526, which may drain into an outlet, like coolant supply line 345 (FIG. 3) or piping 411 (FIG. 4), for example. In the instance of coolant supply line 345 in an example embodiment filter 300 of FIG. 3, suction from an induction pump may be sufficiently large to overcome pressure drop across each layer, driving and filtering coolant through filter 500 in sufficient volumes to create a larger, combined and clean induced flow of at least 2 m/s. In this way an induced flow may not only reduce radionuclide particulate depositions on surfaces immersed in coolant, but it may also propel a mobile assembly cleaning the same and filter coolant through the mobile assembly near an area likely to have much coolant contaminate to be intercepted through example embodiment filters. Example embodiment filter 500 may be constructed in a manner that permits easy assembly/disassembly and minimizes additional handling of potentially radioactive components post-use. For example, each stage 534, 533, 532, 531, and 530 may be contained in a resilient filter segment with exterior flanges 501 around a perimeter of each segment end. Each flange 501 may seal against an adjacent flange between adjacent segments with a quick release 502 like a buckle or fastener that allows individual segments to be easily removed for cleaning and/or disposal at flanges 501. Flanges 501 and releases 502 may be compatible with high integrity disposal systems in shape and joining structure to permit direct disposal of used, dirty filter elements from a filter segment. Further, flanges 501 may accommodate additional shielding and/or flotation rings to be added to filter 500. For example, a dense shielding ring, such as one made out of tungsten, may be added to surround filter 500 and sit against flanges 501 to minimize exposure during handling. Similarly, a buoyant floatation ring may pass around a segment of filter 500 under a flange 501 and change buoyancy of filter 500 and example embodiment mobile assembly 300 (FIG. 3) to allow desired buoyancy and movement in coolant. Example embodiment system 100, including a base 120 and/or mobile assembly 150 and their example embodiment components 300, 400, 500, may be configured to operate in a nuclear reactor environment. For example, all structural components in example embodiment base 400 and example embodiment mobile assembly 300 may be fabricated of materials designed to substantially maintain their physical characteristics when exposed to radiation, variable temperatures, and caustic environments encountered in nuclear reactors. Similarly, materials used in example embodiments may be of a reliable quality for failure avoidance in probabilistic risk assessment determinations and may be designed to minimize radionuclide particulate or solute entrainment or adsorption to minimize radioactive contamination and cleanup requirements post-use. Example embodiments can be used in a variety of ways to prevent particulate deposition on surfaces immersed in a fluid. For example, in a nuclear power plant, like a BWR, ESBWR, PWR, CANDU, or ABWR, areas, like a refueling cavity or chimney, may be flooded with water coolant during operations and/or maintenance, and example embodiment systems may be installed in such areas to induce coolant water flow of about 2 meters per second against surfaces in the coolant. This may be achieved with an example embodiment mobile assembly creating the flow while immersed in the coolant. Operators may configure and direct example embodiments to specifically position flows about surfaces for deposition removal in the coolant. Example embodiments may also provide active filtering of coolant water in the direct vicinity of the flow that dislodges particulate deposition from the surfaces. Example embodiments may further provide water chemistry with deposition-removing and -dissolving pH, oxidation, replacement cations, etc. By keeping depositions from coolant off of surfaces, radionuclides may not easily remain on submerged surfaces or later become airborne when the surfaces are dried during other operations. Example embodiments and methods thus being described, it will be appreciated by one skilled in the art that example embodiments may be varied and substituted through routine experimentation while still falling within the scope of the following claims. For example, a fluid like light water reactor coolant may be used to create a flow against surfaces in some embodiments, but other fluids, like heavy water, are equally useable in example embodiments. Although example embodiments are shown in parts of a base, mobile assembly, and filter, it is understood that these parts may be combined in a unitary submersible and/or further divided or omitted entirely depending on desired functionality. A variety of different reactor and reactor designs and radwaste management structures are compatible with example embodiments and methods simply through proper dimensioning. All such changes fall within the scope of the following claims, and such variations are not to be regarded as departure from the scope of the following claims. |
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claims | 1. A method of scanning a number of objects in a multi-level environment, comprising:selecting a storage area having a plurality of movable racks, each rack occupying at least one level and capable of carrying at least one object;selecting a scanner;establishing a first vacant multi-level aisle among the racks;moving the scanner substantially vertically within the first vacant aisle to scan successive objects substantially adjacent to the first vacant aisle to detect at least one pre-determined characteristic;establishing at least a second vacant multi-level aisle among the racks; andmoving the scanner substantially vertically within at least the second vacant aisle to scan additional objects substantially adjacent to the second vacant aisle. 2. The method of claim 1 wherein each rack includes a chassis movable along at least one track. 3. The method of claim 1 wherein each rack is movable in at least one side direction. 4. The method of claim 1 wherein each rack is configured to hold at least one intermodal container. 5. The method of claim 4 wherein the scanner is suspended from a crane capable of lifting at least one intermodal container. 6. The method of claim 1 further including generating an alert when the scanner detects the at least one pre-determined characteristic. 7. A method of scanning a number of intermodal containers, comprising:selecting a vessel having at least one cargo hold having a plurality of movable racks, each rack occupying at least one level and capable of carrying at least one container;selecting a scanner;establishing a first vacant multi-level aisle among the racks;moving the scanner substantially vertically within the first vacant aisle to scan successive containers substantially adjacent to the first vacant aisle to detect at least one pre-determined characteristic;establishing at least a second vacant multi-level aisle among the racks; andmoving the scanner substantially vertically within at least the second vacant aisle to scan additional containers substantially adjacent to the second vacant aisle. 8. The method of claim 7 wherein each rack includes a chassis movable along at least one track secured within the cargo hold. 9. The method of claim 7 wherein each rack is movable in at least one side direction. 10. The method of claim 7 wherein the scanner is suspended from a crane capable of lifting at least one intermodal container. 11. The method of claim 7 further including generating an alert when the scanner detects the at least one pre-determined characteristic in a container. 12. The method of claim 11 further including placing containers having the detected pre-determined characteristic in at least one of quarantine and disposal. 13. The method of claim 7 wherein at least two sides of substantially each container are scanned. 14. The method of claim 7 wherein the containers are scanned while the vessel is underway. 15. A system of scanning a number of intermodal containers, comprising:a storage area having a plurality of movable racks, each rack occupying at least one level and capable of carrying at least one container, the racks capable of being moved to establish successive vacant multi-level aisles among the racks; andat least one scanner movable substantially vertically within the successive vacant aisles to scan successive containers substantially adjacent to each vacant aisle to detect at least one pre-determined characteristic. 16. The system of claim 15 wherein each rack includes a chassis movable along at least one track. 17. The system of claim 15 wherein each rack is movable in at least one side direction. 18. The system of claim 15 wherein the scanner is suspended from a crane capable of lifting at least one intermodal container. 19. The system of claim 15 wherein the system generates an alert when the scanner detects the at least one pre-determined characteristic. |
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claims | 1. An imaging system, comprising:an electromagnetic interference (EMI) shield configured to shield one or more imaging components from electromagnetic interference, wherein the EMI shield comprises a first material having a first plurality of conductive elements integrally formed within a first nonconductive material, wherein the first material has a first generally nonconductive exterior; anda conductive member having first and second portions, wherein the first portion extends into a first recess into the first material through the first generally nonconductive exterior in contact with the first plurality of conductive elements, and the second portion protrudes outside of the first generally nonconductive exterior. 2. The imaging system of claim 1, wherein the EMI shield surrounds an image detector panel. 3. The imaging system of claim 1, wherein the EMI shield at least substantially defines a panel-shaped portable housing. 4. The imaging system of claim 1, comprising an imaging component shielded by the EMI shield. 5. The imaging system of claim 4, wherein the imaging component comprises an x-ray component. 6. The imaging system of claim 1, wherein first plurality of conductive elements comprise fibers, particles, or a combination thereof. 7. The imaging system of claim 1, wherein the first material comprises a composite material, a compounded plastic, or a combination thereof. 8. The imaging system of claim 1, wherein the first plurality of conductive elements comprises stainless steel fibers and the first nonconductive material comprises polycarbonate. 9. The imaging system of claim 1, wherein the first plurality of conductive elements comprises carbon particles, or fibers, or a combination thereof, and the first nonconductive material comprises polycarbonate. 10. The imaging system of claim 1, wherein the EMI shield comprises a first component made of the first material and a second component made of a second material different from the first material, wherein the second material comprises a second plurality of conductive elements integrally formed within a second nonconductive material, and the second material has a second generally nonconductive exterior, wherein the second portion of the conductive member extends into a second recess into the second material through the second generally nonconductive exterior in contact with the second plurality of conductive elements. 11. The imaging system of claim 1, comprising a secondary shielding layer. 12. The imaging system of claim 11, wherein the secondary shielding layer comprises a conductive paint, a metallic foil, a woven fabric, or a combination thereof. 13. A method for shielding electromagnetic interference in an imaging system, comprising:providing an electromagnetic interference (EMI) shielding enclosure comprising a first material consisting essentially of a first plurality of conductive elements disposed in a first non-conductive material and a second material consisting essentially of a second plurality of conductive elements disposed in a second non-conductive material; andconductively coupling the first material with the second material to form a conduction path between the first plurality of conductive elements and the second plurality of conductive, wherein conductively coupling comprises abrading a non-conductive surface of the first material, or the second material, or both, to reveal a conductive surface having at least some of the conductive elements exposed. 14. The method of claim 13, wherein conductively coupling the first material with the second material comprises extending a conductive interface structure into a recess in the first material or the second material. 15. The method of claim 14, wherein extending the conductive interface structure comprises inserting or overmolding a metal stud in the first material or the second material. 16. The method of claim 13, wherein the first material, or the second material, or both comprise a compounded plastic. 17. The method of claim 13, wherein the first material, or the second material, or both comprise a composite material. 18. An imaging system, comprising:image detection circuitry; anda portable enclosure disposed about the image detection circuitry, wherein the portable enclosure comprises;a first component at least substantially made of a first electromagnetic interference (EMI) shielding material, wherein the first EMI shielding material comprises a first plurality of conductive elements disposed in a first non-conductive material, and the first component comprises a first non-conductive surface disposed over the first EMI shielding material;a second component at least substantially made of a second electromagnetic interference (EMI) shielding material, wherein the second EMI shielding material comprises a second plurality of conductive elements disposed in a second non-conductive material, the second component comprises a second non-conductive surface disposed over the second EMI shielding material, the first and second components are coupled together along an interface, and the first and second plurality of conductive elements are conductively coupled together via a conduction path through the first and second non-conductive surfaces along the interface. 19. The system of claim 18, wherein the conductive path comprises a conductive stud overmolded in the first EMI shielding material, or second EMI shielding material, or both. 20. The system of claim 18, wherein the conductive path comprises an abraded surface of the first EMI shielding material, or second EMI shielding material, or both. 21. The system of claim 18, wherein the image detection circuitry comprises an x-ray detector panel. 22. The system of claim 18, wherein the portable enclosure has a panel-shaped geometry. 23. The system of claim 18, wherein the first non-conductive material, or the second non-conductive material, or both, comprises polycarbonate, and the first plurality of conductive elements, or the second plurality of conductive elements, or both, comprises carbon fibers, or carbon powder, or stainless steel fibers, or a combination thereof. 24. The system of claim 18, wherein the first EMI shielding material or the second EMI shielding material is a compounded plastic. 25. The system of claim 18, wherein the first EMI shielding material or the second EMI shielding material is a composite material. 26. A method for shielding electromagnetic interference (EMI) in an imaging system, comprising:providing an EMI shielding enclosure comprising a first material having a non-conductive surface, wherein a second EMI shielding material is disposed on the non-conductive surface of the first material; andelectroplating or electroless plating the EMI shielding material onto the non-conductive surface of the first material. 27. The method of claim 26, comprising painting the second EMI shielding material onto the non-conductive surface of the first material. 28. The method of claim 26, wherein the second material comprises a metallic foil. 29. The method of claim 26, wherein the second material comprises a woven fabric. |
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summary | ||
claims | 1. A control rod/control rod drive mechanism (CRDM) coupling comprising:a connecting rod operatively connected with a CRDM unit to provide at least one of gray rod control functionality and shutdown rod control functionality; anda terminal element connected with a lower end of the connecting rod, the terminal element including a casing defining at least one cavity and a filler disposed in the at least one cavity, the filler comprising heavy material having a higher density than a material comprising the casing, the terminal element further connected with an upper end of at least one control rod. 2. The control rod/CRDM coupling as set forth in claim 1, wherein the material comprising the casing is stainless steel. 3. The control rod/CRDM coupling as set forth in claim 2, wherein the terminal element has an average density that is greater than the density of stainless steel. 4. The control rod/CRDM coupling as set forth in claim 2, wherein the heavy material has a density of at least 16.2 grams per cubic centimeter at room temperature. 5. The control rod/CRDM coupling as set forth in claim 2, wherein the heavy material is selected from a group consisting of tungsten, depleted uranium, molybdenum, and tantalum. 6. The control rod/CRDM coupling as set forth in claim 2, wherein the filler comprises heavy material in the form of a powder or granulation. 7. The control rod/CRDM coupling as set forth in claim 1, wherein the heavy material has a density that is at least twice the density of the material comprising the casing. 8. The control rod/CRDM coupling as set forth in claim 1, wherein the terminal element further includes one or more casing cover plates that seal the at least one cavity of the casing. 9. The control rod/CRDM coupling as set forth in claim 1, wherein the terminal element has elongation in a SCRAM direction that is at least as large as a largest dimension of the terminal element transverse to the SCRAM direction. 10. The control rod/CRDM coupling as set forth in claim 1, wherein a cross-section of the terminal element oriented broadside to a SCRAM direction has an area fill factor of at least 50%. 11. The control rod/CRDM coupling as set forth in claim 1, wherein the connecting rod comprises a hollow connecting rod tube and a filler disposed in the hollow connecting rod tube, the filler comprising heavy material having a higher density than a material comprising the hollow connecting rod tube. 12. The control rod/CRDM coupling as set forth in claim 1, further comprising a J-Lock coupling connecting the terminal element with the lower end of the connecting rod. 13. An apparatus comprising:a nuclear reactor pressure vessel; anda control rod assembly including at least one movable control rod comprising a neutron absorbing material, a control rod drive mechanism (CRDM) for controlling movement of the at least one control rod, and a coupling operatively connecting the at least one control rod and the CRDM, the coupling including a connecting rod engaged with the CRDM and a terminal element connected with a lower end of the connecting rod, the terminal element including a first portion comprising a first material having a first density and a second portion comprising a second material having a second density that is greater than the first density, the terminal element further connecting with the at least one control rod, wherein the connecting rod is detachably engaged with the CRDM such that detachment of the detachable engagement causes a translating assembly including at least the connecting rod, the terminal element, and the at least one control rod to fall toward a reactor core disposed in a lower region of the nuclear reactor pressure vessel, and wherein the first portion of the terminal element supports or contains the second portion of the terminal element. 14. An apparatus comprising:a nuclear reactor pressure vessel; anda control rod assembly including at least one movable control rod comprising a neutron absorbing material, a control rod drive mechanism (CRDM) for controlling movement of the at least one control rod, and a coupling operatively connecting the at least one control rod and the CRDM, the coupling including a connecting rod engaged with the CRDM and a terminal element connected with a lower end of the connecting rod, the terminal element including a first portion comprising a first material having a first density and a second portion comprising a second material having a second density that is greater than the first density, the terminal element further connecting with the at least one control rod, wherein the first portion of the terminal element comprises a steel enclosure enclosing the second portion of the terminal element. 15. The apparatus as set forth in claim 14, wherein the second density is at least 16.2 grams per cubic centimeter at room temperature. |
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039327485 | abstract | Method of determining the distance between an area under fire and the muzzle of the weapon by means of autoradiography of the irradiated firing residues on this area. Several carrier areas are fired at from various distances. The carrier areas are subsequently activated by neutron irradiation and subsequently contacted with a film sensitive to nuclear radiation. The series of autoradiographs produced after development on the film are compared with the autoradiograph, produced in the same way, of the firing trace to be investigated on the area fired at and used as a distance standard for that trace. |
051209730 | description | DETAILED DESCRIPTION OF PREFERRED EMBODIMENTS OF THE INVENTION In its simplest form the device according to the invention consists of a shielded loading and/or storage container 4 for radiation sources, which are not shown in detail. This container 4 is connected via a channel 16 with a switch 17 that can be operated by a drive motor 18. A switch of this kind is described in more detail in European Patent No. 0 128 300. From the switch 17 a further channel 9 leads via a coupling 10 and a connection plate 11 to an applicator 12. The side of the switch 17 remote from the channels 9 and 16 is connected, via a rotatable channel 22, a short channel section 43 and a release means 21, to a fork 20, the channels 23, 24 of which lead to respective drives 25, 26. Such drives are described in more detail, for example in German patent specification 33 35 438. By means of the drives 25, 26 thrust and traction wires (not shown}can be passed through the channels 23, 24, 43 and the rotatable channel 22 either via the channel 16 into the loading and/or storage holder 4 or via the channel 9 to an applicator 12. As shown in FIG. 2, the traction wire 33 moved by the drive 25 has at its end a bush 34 of slightly greater diameter than the traction wire 33 and continues as a pin 35 of smaller diameter. A needle-shaped holder 40 (FIG. 4) filled with radioactive material forms the radiation source that is to be brought from the loading and/or storage container 4 and inserted into the applicator 12. At the end of the needle-shaped holder 40 there is a sleeve 41 with inwardly-facing spring elements 42. The external diameter of the pin 35 is adapted to the internal diameter of the sleeve 41 so that on pushing the pin 35 into the sleeve 41 a clamping connection is formed. If the traction wire 33 is now advanced into the loading and/or storage container 4 by means of the drive 25, with the switch 17 appropriately set, the pin is pushed into the sleeve 41, since the needle-shaped holder 40 is in its end position. If the drive 25 is operated in the opposite direction the needle-shaped holder 40 can be withdrawn by means of the traction wire 33 until it reaches the release means 21. The release means 21 consists in its simplest form of a constriction before the fork 20 through which the cable 33 with the bush 34 and the pin 35 can pass, while the diameter of the sleeve 41 is so great that it strikes against the release 21. At this point the needle-shaped holder 40 is therefore caused to separate from the traction wire 33. The drive 26 is then switched on and the appropriate thrust wire 36(FIG. 3) pushes against the sleeve 41 with its bush 37 and a cylindrical extension 38. In the same way, the cylindrical extension 38 can carry a pin 39 with a diameter small enough for it not to be gripped by the inwardly-facing spring elements 42 of the sleeve 41 so as to avoid a coupling efffective in the direction of traction being reformed. By means of the thrust wire 36 the needle-shaped holder 40 can be moved into the applicator 12 after changing over the switch 17 with the rotatable channel 22. If the thrust wire 36 is now moved back into the starting position, the needle-shaped holder 40 remains in the applicator 12, so that the applicator coupling 10 can be released without difficulty, thus giving the patient with the applicator 12 considerably more freedom of movement. In FIG. 5 a form-locking coupling, effective in the direction of traction, is shown together with the associated release means 21. In this case the needle-shaped holder 40 has a slit sleeve 44, with resilient hooks 45 arranged in the slits. The resilient hooks cooperate with a locking groove 48 in a pin 47 at the end of a traction wire 46. FIG. 5 shows that the pin 47 can be pushed into the sleeve 44 between the resilient hooks 45, but after engagement of the resilient hooks 45 in the groove 48 is firmly connected to the traction wire 46 in the direction of traction. To release the needle-shaped holder 40 from the traction wire 46 the traction wire 46 is withdrawn far enough for oblique faces on the resilient hooks 45 reach the region of wedge surfaces 49 on the release 21. The resilient hooks 45 are thereby opened out and come out of the groove 48, and the traction wire can be released from the needle-shaped holder 40. A channel 43 on the release 21 provides sufficient space for the resilient hooks 45 to move apart in the region of the release 21, but elsewhere ensures precise guidance of the needle-shaped holder 40. FIGS. 6 and 7 illustrate the device according to the invention in detail. A housing 3 is mounted on a carriage 1 with rollers 2, leaving room on the carriage 1 for a shielded loading and/or storage container 4. The loading and/or storage container 4 is connected to the housing 34 through a tube 5. The loading and/or storage container 4 may, for example, have 16 storage channels 7 for radiation sources in the form of the needle-shaped holders 40 shown in FIG. 4. Inside the tube 5 are sixteen corresponding wire channels 6 running to the switch 17. From the same side of the switch 17 channels 15 lead to a connection plate 8, likewise for sixteen wire channels 9. These wire channels 9 end in the applicator coupling 10 which can be coupled to the connection plate 11 for the applicator channels. Sixteen applicators 12 in the form of needles are connected to this connection plate 11. The coupling 10 includes optical means for monitoring proper connection of the connection plate 11 with the coupling 10, in the form of an optical fibre 13 which runs from the coupling 10 to an optical fibre connection 14. This prevents thrust or traction wires and/or radiation sources from being moved out of the coupling 10 when this is not connected with the connection plate 11. In order to position the rotatable channels 22 precisely on the switch 17 for the connections for the channels 6 and 15, a catch 19 engages in the switch 17 in each position to which it is moved. Before each rotation of the switch, a test is carried out by means of a mechanical probe to ensure that there is no wire or radiation container in the switch. Since the wire channels to the patient being treated can be from one to three meters long, the thrust or traction wires moved by the drives 25, 26 must also be correspondingly long, so storage channels 27, 28 are provided on the back of the drives 25, 26 into which the thrust or traction wires can be passed until they come up against limit switches 29, 30 that switch off the drives 25, 26. The device is constructed in the form of a desk and has a front plate 31 carrying LCD indicators and buttons. A container 32 with an accumulator and an electronic control unit are housed in the lower part of the housing 3. By means of the device according to the invention radiation sources can be inserted into and withdrawn from applicators in a programmed manner, without the operators being exposed to radioactive radiation. |
abstract | A device serves for controlling temperature of an optical element provided in vacuum atmosphere. The device has a cooling apparatus having a radiational cooling part, arranged apart from the optical element, for cooling the optical element by radiation heat transfer. A controller serves for controlling temperature of the radiational cooling part. Further, the device comprises a heating part for heating the optical element. The heating part is connected to the controller for controlling the temperature of the heating part. The resulting device for controlling temperature in particular can be used with an optical element in a EUV microlithography tool leading to a stable performance of its optics. |
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description | This invention pertains to a protective garment such as for a firefighter or emergency worker, and in more particular applications, to a protective garment having including a drag harness. Commonly, a firefighter or an emergency worker wears a protective garment, such as a protective coat. Furthermore, firefighters or emergency workers also wear additional safety equipment, such as drag harnesses, such that the wearer can be dragged and/or carried by a rescuer should the wearer become incapacitated. These drag harnesses can be worn within or on the exterior of the of the protective garment. Protective garments and drag harnesses have generally been configured to have a pull strap located behind the head of the wearer. In this form, the pull strap can be extended from the protective garment to drag the wearer should the wearer become injured or incapacitated. However, the drag harness is susceptible to catching on or becoming abraded by external surfaces. Therefore, drag harnesses may be located completely within the protective garment and/or covered by pockets or flaps. However, the drag harness must still be capable of being grasped quickly in an emergency situation. Furthermore, the rescuer is generally wearing bulky gloves which provide diminished tactile feedback. In one form, a protective garment for a firefighter or emergency worker is provided. The protective garment includes an outer shell and a drag harness. The drag harness is located at least substantially within the outer shell and includes a wearer portion, a gripping portion and a flap operably coupled to the gripping portion. The flap is releasably secured to the outer shell in a stored state. The flap and gripping portion remain operably coupled to one another and extend away from the outer shell in an deployed state to drag a wearer. According to one form, a protective garment for a firefighter or emergency worker is provided. The protective garment includes an outer shell and a drag harness. The outer shell has at least one aperture. The drag harness is located at least substantially within the outer shell. The drag harness includes a wearer portion, a gripping portion and a flap operably coupled to the gripping portion. The flap releasably is secured to and substantially covers the at least one aperture in a stored state. The flap is graspable by a rescuer to pull the flap away from the outer shell and the gripping portion through the aperture in a deployed state to drag a wearer. In one form, wherein the outer shell includes two apertures. According to one form, at least a part of the gripping portion is permanently affixed to the flap. In one form, the flap comprises an outer surface and an interior space such that a part of the gripping portion passes through the interior space and the flap is movable along the gripping portion. According to one form, the flap is a gripping handle whereby a rescuer may drag the wearer. In one form, the protective garment further includes at least one releasable fastener selected from the group comprising hook and loop fasteners, snap fasteners and button fasteners to releasably secure the flap to the outer shell in the stored state. According to one form, the outer shell is made from a fire resistant material and the flap is made from the same fire resistant material. In one form, substantially all of the outer shell has an outer appearance of a first visible color and the flap has an outer appearance of a second visible color which contrasts the first color. Other objects, features, and advantages of the invention will become apparent from a review of the entire specification, including the appended claims and drawings. As illustrated in FIG. 1, a protective garment 20 is shown. The protective garment may take a variety of forms such as a protective coat 22 or protective pants (not shown). The protective garment 20 may be similar to many conventional types of protective garments known to those skilled in the art and therefore those common features will not be discussed in detail herein. For example, the protective garment may include a protective outer shell and one or more thermal and/or water resistant liners. As described herein the protective garment 20 includes additional features which will be detailed below. Furthermore, it should be understood that these additional features may be added to many forms of existing protective garments such that the garment may be retro-fit to accommodate the additional features. The protective garment 20 includes an outer shell 24 having at least one aperture 26. However, it should be understood that the garment 20 may include multiple apertures 26, such as seen in FIGS. 3 and 4. The aperture(s) 26 can be used to provide an opening to gain access to various components or objects located within the outer shell 24. For example, a drag harness 30 may be located at least substantially within the outer shell 24. Referring to FIGS. 1 and 3, in one form the drag harness 30 includes a wearer portion 32 and a gripping portion 34. The wearer portion 32 extends at least partially around a portion of the wearer. For example, in the form shown in FIG. 3, the wearer portion 32 extends at least partially around the torso 40 of the wearer. In other forms, the wearer portion may extend at least partially around a limb of the wearer, such as an arm or leg. Furthermore, the drag harness 30 may include multiple wearer portions 32. The gripping portion 34 may be a separate component affixed to the wearer portion 32, or may be an integrated as part of the wearer portion 32 to form a single loop, as shown in FIG. 3. Generally, the gripping portion 34 can be used by a rescuer to grip the drag harness 30 to drag the wearer. In this manner, when the gripping portion 34 is pulled by the rescuer, the wearer portion 32 will tighten against at least a portion of the wearer so that the wearer can be dragged and/or carried. As shown in FIG. 1, one form of a flap 50 is shown. In this form, the flap 50 includes an outer surface 52 and at least one releasable fastener 54. As seen in FIG. 1, the releasable fastener 54 includes hook 56 and loop 58 portions to releasably secure the flap to the protective garment 20. However, it should be understood that other types of fasteners may be used such as a snap 60 as found in FIG. 6, a button 62 as found in FIG. 7 and other forms of releasable fasteners as understood by those skilled in the art. Furthermore, referring again to FIG. 1, the releasable fastener 54 is shown surrounding the aperture 26. It should be understood that this form may be used with multiple apertures 26, such as shown in FIG. 3. In another form, as shown in FIG. 4, the releasable fastener 54 does not surround the apertures 26, but instead is located adjacent a portion of the apertures 26. It should be understood that this form may also be used with a single aperture 26. Furthermore, it should be understood that the releasable fastener 54 may be located adjacent other portions of the aperture 26 and/or may be located remotely from the apertures 26. In one preferred form, the flap 50 is sized such that it substantially covers the one or more apertures 26. In this regard, the flap 50 can prevent moisture and debris from entering the aperture 26. Furthermore, the size and orientation of the flap 50 can prevent the drag harness 30 from being snagged and/or abraded on external surfaces. The flap 50 may be operably coupled to the drag harness 30 in a variety of manners. For example, referring to FIG. 1, the flap 50 is operably coupled to the gripping portion 34 of the drag harness 30 such as by sewing the flap 50 to the gripping portion 34. In this form, the flap 50 can help prevent the drag harness 30 from shifting significantly as it is worn by a wearer. In the embodiment of FIG. 1, the gripping portion 34 is a continuous length of material. It should be understood that this form may also be implemented wherein the gripping portion 34 includes two ends which are each sewn to the flap 50. Furthermore, it should be understood that other methods may be utilized to connect the flap 50 to the gripping portion 34, such as rivets, adhesive and other forms understood by those skilled in the art. Referring to FIG. 4, the flap 50 is operably coupled to the gripping portion 34 in another manner. In this form, the flap 50 includes an interior space 70 whereby the gripping portion 34 is permitted to pass through. As shown in FIG. 4, the interior space 70 is accessible via apertures 72 in the outer surface 52. However, it should be understood that the apertures 72 may also be located at other locations on the flap 50, such as at ends 74. As the gripping portion 34 extends through the interior space 70 and is not secured thereto, the flap 50 is permitted to move along the length of the gripping portion 34, such as when a rescuer is dragging the wearer. However, it should be understood that the gripping portion 34 may be secured to the flap if desired. It should be understood by those skilled in the art that the forms illustrated in the figures may be intermixed to produce a desired combination of elements. For example, the number of apertures 26, the number and orientation of the releasable fasteners 54, the method of coupling the flap 50 to the gripping portion 34, as well as the type of flap 50 may be intermixed to produce a desired combination of elements. It should also be understood that the drag harness 30 may also be utilized in other protective garments, such as protective pants. In this form, the protective garment would include the outer shell 24 and the drag harness 30 would still include the wearer portion 32, the gripping portion 34 and the flap 50. Again, any combination of elements may be chosen to produce a desired combination of elements. It should be understood that the drag harness 30 may be sewn into a layer of the protective garment 20. In another form, the drag harness 30 may also be releasably secured within the protective garment 20 by a variety of fastening means known by those skilled in the art, such as snaps, hook and loop fasteners and the like. The protective garment 20, drag harness 30 and flap 50 may be made from a variety of materials. Furthermore the protective garment 20, drag harness 30 and flap 50 may be made of the same or different materials. In one form, the drag harness 30 and the flap 50 are made of fire resistant material, such as Nomex® or Kevlar®. However, it should be understood that a variety of other materials may be used. Furthermore, the drag harness 30 may be made of a rope-type material, a web-type material and other forms understood by those skilled in the art. The drag harness 30 may be made of different materials based upon the location of the drag harness 30 on the wearer's body. For example, the gripping portion 34 may be made of different materials than the wearer portion 32. The flap 50 can be used as a gripping handle for grasping the gripping portion 34 of the drag harness 30. In this regard, the flap 50 can include reinforcing structure to make the flap more rigid and potentially easier to grasp. Furthermore, the flap 50 can be made from a material having a specific color. For example, the outer shell 24 may be a dark color while the flap 50 can be a lighter color and/or made from a reflective material to increase the visibility of the flap 50. The flap 50 and the gripping portion 34 can be used by a rescuer to drag and/or carry the wearer. The flap 50 is releasably secured to the outer shell 24 in a stored state, but may be removed to extend away from the outer shell 24 in a deployed state to drag the wearer. A rescuer can pull on the flap 50, which is completely separable from the outer shell 24 and can grasp the flap 50 and/or gripping portion 34. It should be appreciated that for all of the disclosed embodiments there are many possible modifications. Additionally, it should be understood that the embodiments described herein may be utilized in conjunction with one another or separately. |
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052001172 | abstract | Alkaline earth metal scales, especially barium sulfate scale deposits are removed from oilfield pipe and other tubular goods with a scale-removing composition comprising an aqueous alkaline solution having a pH of about 8 to about 14, preferably about 11 to 13, of a polyaminopolycarboxylic acid, preferably EDTA or DTPA and a catalyst or synergist comprising a monocarboxylic acid, preferably a substituted acetic acid such as mercaptoacetic, hydroxyacetic acid or aminoacetic acid or an aromatic acid such as salicylic acid. When the scale-removing solution is contacted with a surface containing a scale deposit, substantially more scale is dissolved at a faster rate than is possible without the synergist. |
047675942 | summary | BACKGROUND OF THE INVENTION This invention relates to sodium cooled reactors. More particularly, this invention relates to an improved reactor vessel auxiliary cooling system sodium flow circuit to supplement heat discharge through the reactor vessel to passing air for residual heat removal from a sodium reactor shutdown under emergency conditions. OUTLINE OF THE DISCLOSURE In certain sodium cooled reactors, the reactor vessel and containment vessel have immediate their exterior an air cooling system. This air cooling system provides for the dissipation of residual heat upon emergency shutdown of the reactor. Since such air cooling systems are well known in the prior art, they will not be discussed further here. This invention is directed rather to the dissipation of heat through the reactor vessel and containment vessel walls where it may reach the air cooling system. In certain sodium cooled reactors, the sodium hot pool is separated from the sodium cold pool by a reactor vessel liner. The purpose of the reactor vessel liner is to separate the reactor hot pool from the reactor cold pool and force fluid flow through the intermediate heat exchanger (IHX) located within the reactor vessel. The reactor vessel liner has a vital secondary function. That function is to short circuit the flow through the IHX to the reactor vessel liner flow gap immediate the reactor vessel wall. Such a short circuiting is required for residual heat dissipation upon loss of normal heat removal systems. The residual heat escapes through the reactor vessel and containment vessel. In such a casualty, it is assumed that reactor control rods are fully inserted. With such full insertion, there nevertheless remains residual heat that must be dissipated. It is the dissipation of this residual heat and the activation of the coolant flow path (here liquid sodium) which is the subject of this invention. In the understanding of this invention, extensive attention will be directed to the prior art normal operation flow path and the prior art residual heat discharge flow path. Emphasis will be placed upon the shortcomings of the prior art residual heat discharge flow path. Thereafter, and once these shortcomings are understood, the improvement constituting the addition of jet pumps from the cold pool to discharge at the slightly higher pressure hot pool will be set forth. It will be emphasized that an improved safety circuit is disclosed. |
abstract | The present invention provides an operation method of a plant which has a low-pressure feed water heater, a deaerator and a high-pressure feed water heater sequentially arranged in a feed water pipe reaching a steam generator from a condenser, and leads high-temperature feed water to the steam generator, wherein an oxidant is injected onto a surface of a structural material from an oxidant injection line in order to form a film that suppresses an elution of an element constituting the structural material such as the feed water pipe, the low-pressure feed water heater, the deaerator and the high-pressure feed water heater, which come in contact with the high-temperature feed water, and a corrosion suppression substance is further introduced from a corrosion suppression substance introduction line in order to deposit the corrosion suppression substance on a surface of the structural material in which corrosion accelerated by a flow of the feed water occurs. |
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055925200 | claims | 1. A control rod for a nuclear reactor comprising: a control rod body having a plurality of blades projecting generally at right angles to one another, said body having a window defined by a plurality of generally linearly extending sides; a latch handle for connection to said control rod body and location in said window, said latch handle having a plurality of linearly extending sides, at least a pair of sides of one of said window and said latch handle having retaining slots and at least a pair of sides of another of said window and said latch handle having flanges for engaging in said slots; the sides of said latch handle and said window being configured so that said latch handle, in a first rotational orientation relative to said window, is receivable within the peripheral confines of said window and, upon rotation thereof into a second rotational orientation relative to said window, engages the periphery of said window with slots and flanges of said pairs of sides of said window and said latch handle engaging one another, respectively, to retain said latch handle in said window; said pairs of sides engaging one another in said second orientation of said latch handle relative to said window to enable sliding movement of said latch handle in at least one linear direction relative to said window. a control rod body having an elongated axis and a plurality of laterally projecting blades, said body having a generally rectangular window defined by upper and lower edges and opposite sides, said window lying along said axis; a generally rectangular latch handle for connection to said control rod body and location in said window, said latch handle having a pair of opposite sides, said opposite sides of one of said window and said latch handle having retaining slots and said opposite sides of another of said window and said latch handle having flanges for engaging in said slots; said latch handle and said window being configured so that said latch handle, in a first rotational orientation relative to said window, is receivable within said window with said latch handle sides lying generally in opposition to said upper and lower edges of said window and, upon rotation thereof into a second rotational orientation relative to said window, has said sides thereof engaging with the opposite sides of said window, respectively, with said slots and flanges engaging one another to retain said latch in said window, said opposite sides of said latch handle and said opposite sides of said window engaging one another in said second orientation of said latch handle to enable sliding movement of said latch handle in at least one linear direction relative to said window and along said axis. a control rod body having a plurality of laterally projecting blades angularly related to one another, said body having a window with spaced edges; a latch handle for connection to said control rod body and location in said window, said latch handle having spaced edges, at least a pair of edges of one of said window and said latch handle having retaining slots and at least a pair of edges of another of said window and said latch handle having flanges for engaging in said slots; the edges of said latch handle and said window being configured so that said latch handle, in a first rotational orientation relative to said window, is receivable within the peripheral confines of said window and, upon rotation thereof into a second rotational orientation relative to said window, engages the periphery of said window with slots and flanges of said edges engaging one another, respectively, to retain said latch handle in said window; said slots and said flanges engaging one another in said second orientation of said latch handle relative to said window to enable sliding movement of said latch handle in at least one linear direction relative to said window. 2. A control rod according to claim 1 wherein said control rod body includes a shaft having an axis and coupled to said latch handle in said second orientation thereof for linear movement therewith in an axial direction of said shaft. 3. A control rod according to claim 2 wherein said latch handle has an opening for receiving said shaft when in said second orientation, a locking device for connecting said shaft and said latch handle to one another in said second orientation of said latch handle relative to said window. 4. A control rod according to claim 1 wherein said latch has a pair of flanges defining said slots on each of said pair of sides of said latch handle, said window having a flange on each of said pair of sides of said window for engaging in said slots. 5. A control rod according to claim 1 wherein said window and said latch handle are generally rectangular in configuration. 6. A control rod according to claim 1 wherein portions of the slots or flanges of one of said pairs of slots and said flanges are arcuate to enable rotation of said latch handle between said first and second rotational orientations. 7. A control rod according to claim 1 wherein said windows and said latch handle are generally rectangular in configuration and wherein portions of the slots or flanges of one of said pairs of slots and said flanges are arcuate to enable rotation of said latch handle between said first and second rotational orientations, the arcuate slots or flanges lying along opposite sides of said latch handle or said window. 8. A control rod for a nuclear reactor comprising: 9. A control rod according to claim 8 wherein said control rod body includes a shaft having an axis and coupled to said latch handle in said second orientation thereof for linear movement therewith in the axial direction of said shaft. 10. A control rod according to claim 9 wherein said latch handle has an opening for receiving said shaft when in said second orientation, a locking device for connecting said shaft and said latch handle to one another in said second orientation of said latch handle relative to said window. 11. A control rod according to claim 8 wherein said opposite sides of said latch handle have set back portions along end portions thereof generally diagonally opposite one another. 12. A control rod according to claim 8 wherein said latch handle has slots along said opposite sides thereof and said window has said flanges along said opposite sides thereof. 13. A control rod according to claim 12 wherein bases of said slots of said latch handle are set back along end portions of said opposite sides generally diagonally opposite one another to enable said latch handle for rotation from said first orientation to said second orientation. 14. A control rod according to claim 13 wherein said control rod body includes a shaft having an axis and coupled to said latch handle in said second orientation thereof for linear movement therewith in an axial direction of said shaft; and wherein said latch handle has an opening for receiving said shaft when in said second orientation, a locking device for connecting said shaft in said latch handle to one another in said second orientation of said latch handle relative to said window. 15. A control rod for a nuclear reactor comprising: 16. A control rod according to claim 15 wherein said control rod body includes a shaft having an axis and coupled to said latch handle in said second orientation thereof for linear movement therewith in an axial direction of said shaft. 17. A control rod according to claim 16 wherein said latch handle has an opening for receiving said shaft when in said second orientation, a locking device for connecting said shaft and said latch handle to one another in said second orientation of said latch handle relative to said window. 18. A control rod according to claim 15 wherein the slots or flanges of one of said pairs of edges and said flanges are arcuate to enable rotation of said latch handle between said first and second rotational orientations. 19. A control rod according to claim 15 wherein said windows and said latch handle are generally rectangular in configuration and wherein the slots or flanges of one of said pairs of edges and said flanges are arcuate to enable rotation of said latch handle between said first and second rotational orientations, the arcuate slots or flanges lying along opposite edges of said latch handle or said window. |
051184613 | claims | 1. A flow measuring apparatus for coolant flow in a boiling-water-reactor which is composed of a pressure vessel; the pressure vessel containing a core, a shroud and a shroud support leg; coolant circulation internal pumps being located in a flow path defined between the outer peripheries of the shroud and shroud support legs and the inner wall of the pressure vessel; and the shroud support leg having a plurality of leg openings, said apparatus comprising: rotational speed detecting means for detecting the rotational speed of the internal pumps; pipe network model calculation means for receiving a rotational speed signal from the rotational speed detecting means and for calculating a flow rate of the coolant at any portion of the flow path on the basis of the rotational speed signal, said model calculation means initially storing a program for a closed pipe network model, corresponding to a flow from a pump outlet of the internal pump the the core, and said model calculation means being responsive to the rotational speed signal to set constants for the model and analytically calculate a flow rate of the coolant at any required portion of that model; and control means for controlling at least one of a coolant flow and a control rod position to control an output power of the reactor, said control means containing said pipe network model calculation means and controlling an output power of the reactor on the basis of a flow signal from the model calculation means. rotational speed detecting means for detecting the rotational speed of the internal pump; pipe network model calculation means for receiving a rotational speed signal from the rotational speed detecting means and for calculating a flow rate of the coolant at any portion of the flow path on the basis of the rotational speed signal, said model calculation means initially storing a program for a closed pipe network model, corresponding to a flow from a pump outlet of the internal pump to a corresponding leg opening, a flow from the pump outlet of the internal pump to a middle point between two adjacent internal pumps in the passage and a flow from the middle point to the leg openings, and said model calculation means being responsive to the rotational speed signal to set constants for the model and analytically calculate a flow rate of the coolant at any require portion of that model; and control means for controlling at least one of a coolant flow and a control rod position to control an output power of the reactor, said control means containing said pipe network model calculation means and controlling an output power of the reactor on the basis of a flow signal from the model calculation means. .alpha..sub.1 : a discharge loss of the nozzle of the internal pump; .alpha..sub.2 : a pressure loss involved from the nozzle section to a middle point between it and two internal pumps in the passage; .alpha..sub.3 : a pressure loss at the location of the leg opening; and .alpha..sub.4 : a pressure loss at the core and the lower plenum. 2. A flow measuring apparatus for coolant flow in a boiling-water-reactor which is composed of a pressure vessel; the pressure vessel containing a core, a shroud and a shroud support leg; coolant circulation internal pumps being located in a flow path defined between the outer peripheries of the shroud and shroud support legs and the inner wall of the pressure vessel; and the shroud support leg having a plurality of leg openings, said apparatus comprising: 3. The flow measuring apparatus according to claim 2, wherein said pipe network model programmed in said pipe network model calculation means is prepared by using .alpha..sub.1, .alpha..sub.2, .alpha..sub.3 and .alpha..sub.4 as flow resistances of corresponding closed pipe network of the model with respective flow/pressure loss characteristics of n number of the internal pumps as differential pressure Pn on those locations of the model corresponding pumps, where 4. The flow measuring apparatus according to claim 2, wherein the pipe network model calculation means comprises a plurality of pump deck differential pressure gauges detecting a differential pressure between suction and discharge sides of the internal pump; speed meters detecting the number of rotations of a respective motor for driving the internal pump; pump section calculation means for receiving a differential pressure signal output from the pump section differential pressure gauge and a rotational speed signal output from the speed transducer, for calculating, from initially measured internal pump characteristics, a discharge of each internal pump on the basis of both the signals, and for calculating a total flow rate of coolant summing up discharges of the pumps; and a pipe network model calculating means, programmed in advance, for receiving a rotational speed signal from each speed transducer, for calculating a flow rate of the coolant from the rotational speed signal, and for delivering a result of calculation as an output so as to adjust and back up the pump section calculation means. |
047675900 | description | DETAILED DESCRIPTION OF THE INVENTION To further understand the present invention, it should be realized that the trajectories of the ions and electrons in a plasma tend very closely to follow or "stick to" the lines of magnetic force. Therefore, it is helpful to describe the character of the lines of magnetic force in a toroidal magnetic confinement device. The "secret" of toroidal magnetic confinement is that the magnetic lines of force in a region of good toroidal magnetic confinement never encounter material objects; therefore plasma particles, whose trajectories tend to follow these lines, do not encounter material objects either. These particles and the plasma they comprise are therefore "magnetically confined" or "magnetically insulated". Explaining the magnetic configuration that gives rise to good toroidal magnetic confinement, one may observe that as the lines of magnetic force in an ideal toroidal magnetic confinement device, such as an ideal tokamak, proceed around the toroid, they actually spiral slowly in a helical manner around one single magnetic line, this special single line being the "magnetic axis". The "magnetic axis" is that line of magnetic force that closes upon itself after going only once around the toroid. Under ideal conditions, the "magnetic axis" would be just a simple circle threading the "center" of the toroidal plasma. Continuing the explanation of the magnetic geometry for toroidal plasma confinement, the volume throughout which a good magnetic confinement configuration can be achieved in any practical device is necessarily limited in spatial extent. The outside surface of this volume is called the "outermost good magnetic surface" and is the imaginary surface generated by the outermost unobstructed line of magnetic force that--after going many many times around the toroid--closes or almost closes upon itself. The volume that lies inside the vacuum vessel but outside the outermost good magnetic surface is called the "scrape-off" region. Obstructions that magnetic lines may encounter in the scrape-off region will include the plasma limiter (i.e., the "scraper") as well as other material objects such as electrodes, coils, diagnostic devices and structural components of the vacuum vessel. Plasma will be present in the scrape-off volume as well as in the volume interior to the outermost good magnetic surface, but the plasma particles in the scrape-off region will relatively quickly strike material obstructions and this portion of the plasma inside the vacuum vessel will not benefit from optimum toroidal magnetic confinement or from optimum magnetic insulation. We will hereinafter refer to the plasma situated interior to the outermost good magnetic surface as the "main plasma" and the magnetic-field-aligned current that flows in the main plasma will hereinafter be referred to as the "main current". And we will hereinafter refer to the vicinity of the outermost good magnetic surface as the "edge" region of the plasma. The edge region will lie principally outside the outermost good magnetic surface in the "scrape-off" region, but it may also be considered to extend somewhat inside this same surface, into the region occupied by the main plasma. Reference will now be made to the preferred embodiment of the present invention, examples of which are illustrated in the accompanying drawings. Referring to FIG. 1, cathode 12 and anode 13 are disposed within the plasma confinement torus 10 and are energized by energizing means 11. Divertor coils 14 are disposed near the cathode 12 and anode 13. Shaping coils 15 are also disposed within the plasma confinement torus 10. The toroidal magnetic field, B , is represented by 16. FIG. 2 shows the calculated poloidal flux contours for a discharge current equal to 1250 amperes for the device of FIG. 1. As shown in FIG. 2, flux contour line 18 is the last unobstructed line of magnetic flux which closes or almost closes upon itself. This condition defines the last good line of magnetic flux and that line, followed many many times around the torus, generates the outermost good magnetic surface. Contour lines 19 are interrupted by a material surface, in this instance, a cathode or anode. The scrape-off region in this illustration is the region outside of contour line 18 and cathode 12 and anode 13 are disposed in the scrape-off region. Now, like plasma particles, plasma current tends strongly to flow along the magnetic lines of force. Thus, current flowing between cathode 12 and anode 13 in FIG. 2 is also disposed in the scrape-off region. The current flowing between cathode 12 and anode 13 will trigger the double-tearing instability only if its direction of flow is parallel (and not anti-parallel) to that of the main current. The region 17 through which this edge current flows is depicted by the shaded region outside the outermost good magnetic surface 18. One criterion for the onset of the double-tearing mode is that the profile of the safety factor, q(r), be double valued in r [that is, that q(r) takes on the same value at two different values of r]. A quick calculation for cylindrical geometry shows that this situation will occur if the density of axial current flow in a cylindrical shell exceeds the average density of axial current flow through the cross-section of the cylinder within that shell. Thus, the double-tearing mode will be driven unstable if the average density of current flow in the shell region is made greater than the average density of current-flow in the main tokamak current channel. It is the double-tearing mode that can provide the anomalous viscosity needed for rapid current penetration. This viscosity can be described mathematically by adding a mean-field term to Ohm's Law. Physically speaking, it is the kinetic momentum of the injected electrons that replaces the momentum lost through electron-ion collisions, and it is the anomalous viscosity that allows this momentum to move inward in minor radius, into the hot interior of the toroidal tokamak plasma. Thus, the plasma current will be augmented by the amount of current which penetrates radially inward from the edge current region 17. FIG. 3 shows another preferred embodiment of the present invention. The plasma 31 is restricted in minor radius by a fractional limiter 33. That is, by a limiter that intercepts only a fraction of the magnetic lines of force in the scrape-off region, just outside the outermost good magnetic surface. Limiter 33 is here made the anode of an electrode system, and one or mor cathodes 32 are placed in the scrape-off region, close to the outermost good magnetic surface. The cathodes 32 are disposed so that they can emit electrons only in the direction parallel to the direction of average electron flow in the main current channel. After leaving a cathode 32, electrons follow along the lines of magnetic force 35, spiral slowly around the magnetic axis 36, and circumnavigate the torus several times before hitting the anode 33. In FIG. 3, an imaginary plane bisects the plasma torus and only the more distant half is shown. Successive intersections of the spiralling electron beam are designated by letters a-j. Thus, in the present embodiment, the edge current region forms a toroidal shell that envelopes the toroidal main plasma. A suitable cathode material is lanthanum hexaboride (L.sub.a B.sub.6). FIG. 4 shows another preferred embodiment of the present invention wherein the cathode 43 and the anode 42 are disposed within divertor chamber 48. Divertor coils 45 are arranged such that a divertor x-point is created. This x-point is defined as the point where there is a null in the poloidal maqnetic field. Arrows 46 represent the electron beam path which follows the lines of magnetic force and, again in this embodiment, spirals slowly around the main plasma current channel, circumnavigates the main plasma 41, but eventually re-enters the divertor region and strikes anode 42. In an experiment performed on the ACT-1 toroidal plasma facility, cited in BACKGROUND OF THE INVENTION, it was demonstrated that a low-voltage 200-ampere beam with current density on 100 amperes/cm.sup.2, could be readily achieved. The experiment made use of a lanthanum hexaboride (L.sub.a B.sub.6) cathode which was biased about 300 volts negative with respect to the limiter chamber walls. The current density in a tokamak reactor would have approximately the same value. While Ono et al. recite the use of an electrode system to inject a beam of electrons into a toroidal confinement device, the present invention differs substantially from the ACT-1 experiment described therein. The ACT-1 device was not a tokamak and did not utilize a plasma current to achieve optimum magnetic confinement of the plasma. Ono et al. described the injection of an electron beam directly into a target plasma. The electron beam discharge broke down the background gas to form a highly ionized plasma. Ono et al. further recite that radiofrequency current drive means can be used to drive a steady-state current in the plasma formed by the injected electron beam. Thus, the ACT-1 experiment demonstrated that an electrode system could be used to form a highly ionized plasma but did not show that the electrode system could be used to generate a steady-state current in the plasma. The present invention differs from the experiment described by Ono et al., in that the electrode system would produce an edge region of electric current flow outside of a previously formed, highly ionized plasma. The density of the edge current flow will be sufficiently large, the direction of the edge current will be parallel to that of the main current, and the edge current will be maintained for a sufficient time such that it may penetrate radially into the main plasma by magnetic reconnection or magnetic turbulence. The penetrating current then auguments or maintains the main current of the toroidal plasma. The attraction associated with the preferred embodiments of the present invention described above is that they are totally non-inductive. The electrode system of the above-described preferred embodiments can be energized such that a constant current flows between the cathode and the anode. Alternatively, the electrode system can be pulsed at regular intervals, producing current flow in the scrape-off region parallel to that in the main channel. FIGS. 5(a) and 5(b) show another preferred embodiment of the present invention. Edge current 57 is produced external to plasma 51 by electromagnetic induction. The transformer for producing edge current 57 may comprise coils 53. Coils 53 may themselves comprise toroidal rings disposed around the main plasma torus 50. A conventional tokamak ohmic heating transformer and its associated coils may be used as coils 53 if they are pulsed in the appropriate manner, described below. Alternatively, the transformer for producing the edge current 57 may comprise a ferromagnetic core 52 which surrounds the minor radius of of the main plasma 51 as as depicted in FIG. 5(b). In this disposition, the edge current channel acts as the secondary circuit for the transformer while coils 58 comprise the externally driven primary windings, corresponding to coils 53 in FIG. 5(a). The inductive coils 53 or 58 may be pulsed repeatedly in a manner such that negative and positive increments of current of comparable magnitude are induced on the plasma surface, but only the positive increment is able to turn on the double-tearing instability, which then provides rapid radial penetration for the properly-directed incremental current. By transporting to a region of lower resistivity just the incremental current flowing in one direction, the plasma itself is able to rectify transformer-induce alternating currents. In more detail, consider a tokamak that is already carrying it's equalibrium plasma current, I.sub.o, and consider further that there is no current flowing in the primary circuit of the ohmic heating trasformer. Say then that each new transformer pulse now starts by ramping up the transformer's primary current so that it reaches a value I.sub.1. The transformer induces an incremental current I.sub.2 equal to -I.sub.1 in the plasma, initially on the surface of the plasma torus, which is to flow in a toroidal sense opposite to that of the main plasma current I.sub.o. Say now that the primary current in the transformer, I.sub.1, is then maintained for one or two decay times for the induced incremental plasma surface current. A decay time is here defined as the inductance of the current channel in the plasma divided by its resistance, L/R. After I.sub.2 has undergone appreciable decay, I.sub.1 is driven suddenly back to zero. A new induced counter-current, I.sub.2 '=I.sub.1 appears at the plasma surface. I.sub.2 ' now flows in the same toroidal sense as I.sub.o, and the magnitude of the current density associated with I.sub.2 ' is to be sufficient to drive the double-tearing mode or a similar instability. The instability and associated plasma turbulence tend to flatten the plasma current profile and, in so doing, let some fraction of I.sub.2 ' penetrate rapidly radially inward, to the hotter core region of the plasma that is much less resistive and where the rate of current decay is correspondingly much reduced. The situation is now that the primary current in the transformer is back to its initial value, zero, but the main plasma current, I.sub.o, is increased by the absorbed fraction of I.sub.2 '. The driving force for the unidirectional plasma current is the turbulence-associated viscosity, not the toroidal electric field, and while negative and positive increments of current of comparable magnitude have been induced on the plasma surface (I.sub.2 and I.sub.2 '), rectification occurs because only the positive increment has been able to turn on the instability which offers rapid radial penetration for the incremental current. Referring to the preferred embodiments of the present invention as illustrated in FIGS. 1-4, the electrode system comprising the cathode and the anode may also be pulsed at regular intervals, as described above. Plasma rectification will occur by the same process as that described above for inductive means. The pulse shape and duration must be sufficiently ramped and long, respectively, that instability is not triggered before the counter-current induced on the plasma surface has been largely able to decay away. The final average density of current-flow in the scrape-off region must once more be sufficient to drive the double-tearing mode, which then again provides the mechanism for rapid radial current penetration for some fraction of the incremental current. FIG. 6 illustrates another preferred embodiment of the present invention. A transformer 63 located in the scrape-off region is pulsed to create a current ringlet 65 in the scrape-off region. The ringlet current should flow in the same toroidal direction as the main plasma current. The fundamental poloidal magnetic field in the scrape-off region is shaped to permit the creation of this ringlet, which is itself a "mini-tokamak". Moreover, the magnetic fields of the ringlet transformer 63 should not be linked to the main tokamak plasma 61, and the main plasma current need not be significantly affected by creation of the ringlet 65. As illustrated in FIG. 6, current ringlet 65 is generated near the main plasma 61. Means are provided for pushing the ringlet 65 toward the main plasma. These means may comprise shaping coils 64. The ringlet 65 is pushed toward the main plasma, which process creates an x-point 69, and ringlet 65 is held there while resistive magnetic reconnection takes place, shifting the ringlet current from its original zone to the edge region enveloping the the main tokamak plasma. At this stage, the incremental current is situated in the edge region and forms a shell of current around the plasma 61 similar to the shell of current created by the preferred embodiments illustrated in FIGS. 3, 4 and 5. The discussion of current penetration via double-tearing and/or anomalous viscosity applies then to this case too. In the preferred embodiments discussed up to this point, an incremental current generated in an edge region is caused to penetrate the primary or main current channel of the tokamak, the mechanism for penetration being anomalous viscosity. The incremental current then adds to or augments the main current. Alternatively, the main current itelf may be started up ab initio by using a cathode and anode as illustrated in FIGS. 7(a) and 7(b). A coaxial pair of toroidal electrodes consisting of cathode 62 and anode 66 is immersed in a magnetic field that is primarily toroidal but which has a small "vertical" component, B.sub.0, directed parallel to the major axis of the system. The electrodes are energized so that electrons, emitted from the cathode, flow in a slowly rising spiral or helicoidal paths that carry them from the cathode to the anode. After switch-on of the electrode circuit, the magnitude of the current flowing between cathode and anode will rise and approach a steady-state value. This steady-state value may even be so large that reversal of the "vertical" magnetic field occurs locally, on one side of the current channel 61. The poloidal projection of the magnetic field lines 65 would then appear as in FIG. 7a. Next, by quickly changing the amount of current being driven through the coils (not shown in FIG. 7) that produce the "vertical" field, the strength of that vertical field, B.sub.0, would itself be changed quickly to a new value, B.sub.0 '. [The change in the strength of the vertical field should be carried out within a time period short compared to the characteristic inductive-resistive (L/R) decay time for the plasma-current "circuit".] The original value of the vertical field, B.sub.0 would have been chosen to optimize the current flow between cathode 62 and anode 66, FIG. 7a. Its new value, B.sub.0 ' should be selected to provide a force equilibrium for the configuration depicted in FIG. 7b, such that the "motor force" between the plasma current and the new verticle field, B.sub.0 ', just balances the self-expansive force, that is, the "hoop force", of the plasma current loop. The external circuit supplying current to the cathode-anode system may, at this point, be switched open (save for voltage-transient protective devices) and the plasma current channel 61 will spontaneously deform itself into iths new equilibrium configuration 63, shown schematically in FIG. 7b. The plasma current channel 63 has now been magnetically detached from the cathode 62 and anode 66 and is closed upon itself, no longer requiring the addition or subtraction of current from external sources such as cathodes or anodes. Region 63, occupied by the main plasma current channel, fills the volume interior to the outermost good magnetic surface 67, and is a region of good magnetic insulation and good magnetic surfaces. It is possible that the processes of anomalous viscosity, magnetic turbulence and magnetic reconnection may be activated by the quick change of vertical field strength described above, and that these "nonclassical" processes will actually hasten the magnetic detachment of the plasma current channel from the material electrodes and enhance the transition to the new equilibrium. It may be understood that alternative means may be used to perform the magnetic detachment of the current channel from the cathodes and anodes. The magnetic field component that performs this detachment may be generated not only by changing the current in the vertical field coils, but also by changing the current in special auxiliary coils that are optimally disposed and programmed to perform the funciton of magnetic detachment of the plasma current channel from the anode-cathode electrode structure. Moreover, during build-up of the electrode current and in performing the transition from the electrode-current phase to the magnetically insulated equilibrium phase, the currents in the vertical field coils and in other coils carrying currents that affect the poloidal magnetic field may all be programmed in time, in direction and in amplitude to effect an orderly build-up and transition and to achieve desired new equilibrium configurations for the magnetic field and for the plasma. Summarizing, an anode-cathode system is so disposed in a mainly toroidal magnetic field that, when these electrodes are energized, a plasma is created and a strong toroidal current is generated. Next, the strength of the "vertical" field is intentionally changed, causing the current channel to seek a new equilibrium configuration. In this process, the current channel carries all or most of the plasma with it as it moves. Provided the vertical field has been programmed correctly, this sequence of events will cause the plasma current channel and all or most of the plasma originally situated in region 61 of anode-cathode plasma current flow, FIG. 7a, to be moved into region 63, FIG. 7b, a region of good magnetic surfaces and good magnetic insulation. The plasma current channel and plasma, now in region 63, are no longer in contact with any material objects. In particular, the plasma current channel and plasma in region 63 are now magnetically insulated from material contact with cathode 62 or with anode 66. This apparatus and method may be used to generate the "start-up" plasma in a conventional toroidal magnetic confinement plasma device such as a "standard" tokamak. The method has the advantages that neither an ohmic heating transformer nor radiofrequency plasma heating nor radiofrequency current drive means are required for its implementation. It should be noted in particular that the use of conventional ohmic heating for "start-up" places severe technological and economic demands on the electrical and structural design of the vacuum vessel and on other apparatus associated with large tokamaks, and the avoidance of this requirement would be considered a strong engineering boon. In addition, it should be mentioned that the startup plasma, after its full creation as depicted in FIG. 7b, could, if desired, be maintained by the methods of steady-state current drive, including the means of anomalous viscosity current drive. For this purpose, the apparatus and methods described in the preferred embodiments of this invention could be used. The disclosed apparatus and method thus provide a system for generating a steady-state toroidal plasma current in a toroidal plasma current in a toroidal confinement plasma device. The disclosed apparatus and method provide a system for generating a steady-state plasma current without using radiofrequency current generating methods. Further, the disclosed apparatus and method provide a system for creating a magnetically confined plasma, ab initio in a toroidal magnetic confinement device. The foregoing description of a preferred embodiment of the invention has been presented for purposes of illustration and description. It is not intended to be exhaustive or to limit the invention to the precise form disclosed, and obviously many modifications and variations are possible in light of the above teaching. The embodiment was chosen and described in order to best explain the principles of the invention and its practical application to thereby enable others skilled in the art to best utilize the invention in various embodiments and with various modifications as are suited to the particular use contemplated. It is intended that the scope of the invention be defined by the claims appended hereto. |
description | This application claims the benefit of priority to U.S. Provisional Application No. 62/460,044, filed Feb. 16, 2017, which is incorporated herein by reference in its entirety. There is disclosed a composition of the wasteform to immobilize both magnox sludges and clinoptilolite wastes. There is also disclosed a method for the conditioning of both the magnox sludge and clinoptilolite type wastes in combination with the additives that enable these wasteforms to be made. Magnox is a type of nuclear power/production reactor that was designed to run on natural uranium with graphite as the moderator and carbon dioxide gas as the heat exchange coolant. The name “Magnox” comes from the magnesium-aluminum alloy used to clad the fuel rods inside the reactor. Magnox, which is short for “magnesium non-oxidizing” has a major disadvantage in that it reacts with water, preventing long-term storage of spent fuel under water. The current baseline method to treat Magnox Sludges and other radioactive sludges is to mix the sludge with a cement grout and cast it inside a container made from steel, such as stainless steel or iron. The cementation route increases the total waste volume, and thus is not ideal. Furthermore, this cemented Magnox sludge is unstable as it contains reactive metal, that leads to hydrogen production and the metal can also react to form expansive phases in the cement product, resulting in swelling of waste container. In addition, the durability of the cemented product is low compared to alternatives such as borosilicate glass or ceramics. HIP technology has been proposed as a method for the immobilization of Magnox sludges and Clinoptilolite zeolite, including co-processing, and minimal process parameters. Vance et. al., Advantages hot isostatically pressed ceramic and glass-ceramic waste forms bring to the immobilization of challenging intermediate- and high level nuclear wastes, Advances in Science and Technology Vol. 73 (2010) pp 130-135). Although the use of Hot-isostatic Press (HIP) technology has been proposed for the treatment of magnox sludge and/or clinoptilolite type wastes previously, at least two essential aspects were missing from these earlier disclosures. The first is the identification of a wasteform chemistry and thus the identification of additives able to safely immobilize the wide range of these wastes after HIPing. In addition, the technical processing challenges associated with treating the inherent hazards associated with these sludges has not been addressed in previous Hot-isostatic pressing (HIPing) disclosures. Namely, there is no disclosure teaching how to deal with flammable metals, or the generation and potential release of unwanted components including hydrogen, carbonates, organics, coarse components, or fine components. The second aspects were missing from these earlier disclosures is a lack of identification of a complete industrial-scale process which can safely and effectively treat the large volumes of these types of waste. To solve the many needs described above, and overcome the mentioned deficiencies, the Inventors have developed compositions and methods to safely condition hazardous sludges and slurries for disposal. While the disclosure describes compositions and methods for magnox sludges and/or clinoptilolite type wastes, as found in magnox reactor sites in the United Kingdowm, the present disclosure is also relevant to other zeolitic ion-exchange materials used in the nuclear industry. Conditioning of the wastes is achieved by the addition of purpose designed additives coupled with the process to passivate the wastes to provide a stable wasteform for subsequent storage and disposal. Therefore, to address the many needs described above, the disclosure relates to the use of special chemical additives in combination with specific process steps, including hot-isostatic pressing as the final consolidation step. There is disclosed a composition to immobilize nuclear containing waste containing at least one radioactive element or alloy of uranium, graphite, magnesium, and aluminum. The composition comprises at least one mineral phase forming element or compound for reacting with the at least one radioactive element or alloy. The composition further comprises at least one glass-forming element or compound to form a glass phase that will incorporate waste radioisotopes and impurities that do not react with the mineral phase forming element or compound. A method of using the disclosed composition to immobilize nuclear containing waste into a solid wasteform is also disclosed. In an embodiment, the method comprises: forming a slurry comprising nuclear containing waste; drying the slurry to form a dried product; calcining the dried product; loading the calcined product into a metal canister; evacuating and sealing the metal canister; and thermally treating the material in the metal canister to produce a dense waste form product. It is to be understood that both the foregoing general description and the following detailed description are exemplary and explanatory only and are not restrictive of the invention, as claimed. The described composition and method can be used to treat hazardous and radioactive sludge, such as settling Pond Sludge including Magnox sludge. “Magnox sludge(s)” is defined as the waste by-product of a Magnox reactor, typically having a wet mud or a similar viscous mixture of liquid and solid components. The waste density is such sludge is typically 1.0-1.4 t/m3, such as 1.2 t/m3 as stored, and 0.6 to 0.8 t/m3, such as 0.7 t/m3 for dry settled sludge. While the disclosed composition and method is described as being particularly beneficial for treating Magnox sludge, in the most basic sense the methods can be used to treat any thick, soft, wet mud or a similar viscous mixture of liquid and solid components, especially containing unwanted radioactive or hazardous materials. One embodiment of the present disclosure is directed to the composition of the wasteform to immobilize waste from magnox reactors, such as magnox sludges and clinoptilolite wastes. The wasteform from such reactors behaves comparatively to high-level waste (HLW) glass wasteforms in aqueous durability testing, while using cheap readily available starting materials. The composition of wasteforms for magnox sludges relate to glass-bonded magnesium silicate/Magnesium titanate matrix. In this case the invention covers addition of calcined magnox sludge at 40 wt %-100 wt % (or the aqueous slurry equivalent) with clinoptilolite (or other zeolitic minerals), silicate minerals, SiO2, TiO2, phosphate, alumina-borosilicate glass, borosilicate glass, silicotitanate glass, iron-phosphate glass, phosphate glass or a combination of these additives. Further additions of lithium oxide, lithium fluoride, calcium fluoride, sodium fluoride, sodium fluorosilicate or other flux, at up to 5 weight % to assist melting may also be used. In addition, titanium, nickel, nickel alloy, iron steel or stainless steel metal powders at up to 10 wt % are also added to control redox conditions in the sludge during HIPing. These are particularly important for the control of the uranium oxide state in the wasteform. Clinoptilolite or other zeolites may also be processed at 100 wt % loading to produce a vitreous wasteform. The waste components are physically diverse and may contain a coarse fraction and a fine fraction of particulates. The relative amounts of each of these size components can vary from 10-90% depending on the source of the sludge. As used herein, “fine” fractions are intended to be submicron, such as less than 200 μm. In one embodiment, fine fractions range from 0.1 to less than 200 μm, such as 0.5 to 150 μm, or even 1.0 to 100 μm. In one embodiment, the waste components comprise fine fractions that contain colloidal precipitates in equilibrium. As used herein, “coarse” fractions are intended to be 200 μm or greater. In one embodiment, coarse fractions range from 200 μm to 6 mm, such as 500 μm to 4 mm, or even 750 μm to 2 mm. In one embodiment, a majority of the coarse fractions have particulate sizes ranging from 200 μm to 6 mm and a majority of the fine fractions have particulate sizes ranging from 0.1 to less than 200 μm. In the present disclosure, there is described a process that changes the targeted phases such that a high waste loading can be achieved with the durability remaining high. For example, in the process the sludge is calcined such that the main component magnesium metal, carbonate or hydroxide is converted to the oxide form or as a component of oxide-minerals. Non-limiting examples of such oxides include:Mg(OH)2→MgO+H2O;MgCO3→MgO+CO2; and2Mg+O2→2MgO. In other embodiments, uranium metal and/or hydrated and carbonated uranium oxides and other significant uranium components are converted to a uranium oxide forms or a component of oxide mineral phases. In various embodiments, the drying and calcination stages remove hydrogen and free water, chemically bound water from the wasteform, making a much more stable product for storage and final disposition than a cementitious route. In a cementitious route hydrogen and hydrogen generating reactive metals remain in the wasteform. Further thermal processing densifies the product to closed porosity (>92% theoretical density) and forms chemical phases and a morphology suitable for long-term storage. In one embodiment, the thermal treatment described herein comprises hot-isostatic pressing. The HIP process is generally described in more detail in U.S. Pat. No. 8,754,282, which is herein incorporated by reference in its entirety. More specifically, as described in this patent, the HIP consists of a pressure vessel surrounding an insulated resistance-heated furnace. Treating radioactive calcine with the HIP involves filling the container with the waste materials, here the contaminated ion exchange media. The container is evacuated and placed into the HIP furnace and the vessel is closed, heated, and pressurized. The pressure is typically provided via argon gas, which, at pressure, also is an efficient conductor of heat. The combination of heat and pressure consolidates and immobilizes the waste into a dense monolith. In an embodiment, the HIP will process one can at a time to a temperature, such as a temperature ranging from of about 800° C. to 1400° C., such as 900° C. to 1300° C. at a processing pressure ranging up to 300 MPa, such as 5 to 150 MPa. The cycle time to process a HIP can ranges up to 16 hours, such as from about 10-16 hours. Once removed from the HIP, the can will be allowed to cool to ambient temperature prior to being loaded into a disposal canister. The HIP temperature may also be modified depending on the waste. Various changes in HIP conditions such as temperatures, pressures, and atmospheres depending on the material being consolidated are discussed in U.S. Pat. Nos. 5,997,273 and 5,139,720, which are herein incorporated by reference. In one embodiment, there is described a method of reacting a small sub <250 um particle sized magnesium-derived component with the additives during the thermal treatment stages calcination and final consolidation. This is achieved by the intermixing of the fine components of the waste with fine additives that provide sources of Ti, Si, P or Al, such as titania, alumina, phosphate, silica and glass frit, to form stable phases. Non-limiting examples of such stable phases for titanate include:MgO+TiO2═MgTiO3 2MgO+TiO2═Mg2TiO4 MgO+2TiO2═MgTi2O5 Non-limiting examples of such stable phases for silicate include:MgO+SiO2═MgSiO3 2MgO+SiO2═Mg2SiO4 A non-limiting example of such stable phases for aluminate includes: MgO+Al2O3═MgAl2O4. In one embodiment, the phases higher in Mg are particularly described phases. In other embodiments, ternary multicomponent phases may also be present, such as from other elements in the waste, including diopside—CaMgSi2O6 and perovskite CaTiO3, formed when calcium is present. The larger sized waste components form MgO grains that are protected by encapsulation in other phase including a silicate glass that is formed from the addition of clinoptilolite, other zeolites, silica, high-silicate minerals or glass frit. The uranium in the sludge is oxidized in the process to form uranium oxide, including UO2 or UO2+x; or reacts with the additives to form a titanate mineral brannerite, pyrochlore, zirconolite. In an embodiment, these may be encapsulated in the matrix. The final waste form is therefore a mixture of MgO and other ceramic phases encapsulated in a protective glass plus ceramic phase matrix. The ability to encapsulate MgO enables high waste loadings (>40 wt %) and protects the MgO from hydration during long-term storage and once disposed in a geological repository. In one embodiment, impurities and fission products present in the waste are incorporated into the phases discussed above or into the glass. Another embodiment of the present disclosure is directed to the process for the conditioning of both the magnox sludge and clinoptilolite type wastes in combination with the additives that enable these wasteforms to be made. In an embodiment, the correct wasteform chemistry and morphology is achieved by processing of the wastes as described below: Slurry mixing: in an embodiment, mixing may utilize paddle type mixing, recirculating mixing, in-line mixing, turbulent slurry mixing or a combination of the above. Drying: in an embodiment, drying is performed using a method whereby granulation occurs concurrently with moisture removal such as wiped or thin-film evaporation, rotary drying or conical mixer/wiper method. However alternative methods may also be utilized including, spray drying, fluidized bed drying or flash drying. Drying also removes any hydrogen present in the waste. Granulation: If required, granulation using roll compaction is utilized after drying to produce a granular product. Calcination: in an embodiment, bound water, carbonates and organics present in the waste can be removed using a calcination methodology such as rotary calcination, vibratory calcination, fluidized bed calcination or a batch calcination method. The calcination step also serves to passivate the reactive metal component of the waste allowing it to react with the additives and become part of a stable wasteform. Loading and packing of a metal hot-isostatic press of hot-press canister/can, with or without vibratory or tamping to improve packing, such as under a dry atmosphere or vacuum. Evacuation and sealing of the metal canister. Thermal treatment to produce a dense waste form product, which may comprise a glass-ceramic enclosed in a metal can. This can be done using either, hot-isostatic pressing, hot pressing (in a bellows or die). It could also be done via sintering of a pellet/puck/block. An example of HIP processing conditions canister occurs at temperatures between 900° C. and 1300° C. and pressures of 5 MPa to 150 MPa. In an embodiment, conditions range from 900° C. and 1050° C., above this an excess of less durable phases form and the glass forming components are consumed to form these less durable phases (magnesium silicates). In one embodiment, one or more additives are provided, as shown in Table 1. TABLE 1Additive Components added at 0-60 wt %Silicate plusComponentSilicate-richTitanate-richtitanateAluminateClinoptilolite or0-1000-5020-600-30similar zeolitemineralTiO2 Titanium0-10 50-95 10-600-10dioxideAlumina0-10 0-550-90 Silica0-1000-3020-500-30Silicate glass frits0-1000-5020-600-50 Compositions In one embodiment, the wasteform compositions are formed during the process and are targeted to increase the durability of the formed wasteform over that of the waste itself or current baseline cementation processes. Additives are mixed with the waste either at the front-end or downstream. These chemical additives react with the waste ions to form target mineral phases in the wasteform structure. All of the compositions contain a glass phase. This glass phase bonds the wasteform and encapsulates the phases. The glass phase is there to incorporate waste radioisotopes and impurities that are not taken up by the mineral phases. In one embodiment, the disclosed method is used to incorporate radioactive caesium isotopes, which along with strontium-90 make up the bulk of the radioactivity and heat generated in waste derived from the nuclear fuel used in power operations. Target mineral phase systems are mainly a combination of titanate, silicate and aluminate phases, plus a silicate glass phase. For high waste loadings the composition contains residual magnesium oxide (MgO) encased in a silicate phase and a silicate glass. In one embodiment, there is described the use of zeolites in the described slurry. The use of Clinoptilolite or other zeolite materials in the waste additive compositions serves at least two purposes. One is to provide the silica needed to make a glass or the silicate phases and the other is to capture any free caesium and other isotopes in the front end and bind them tightly so as to significantly reduce volatile losses during the calcination stage. The described compositions and methods can be used to clean up settling pond sludge surrounding nuclear Decommissioning sites. In these environments, the main source is from fuel corrosion and the nuclides involved include Cs, Eu, Ru, Sr-90 and other mixed fission products and actinides. Uranium is another major radioactive component in Magnox Sludges, which may be immobilized by a method described below. The wastes described herein may be reacted with the inventive composition to form crystalline phases, including pyrochlore, zirconolite and brannerite titanates; uranate phases such as MgUO4, MgUO3.8, MgU3O10. These crystalline phases can then be incorporated into a silicate glass structure, as described herein. In this embodiment, Uranium can be left as residual uranium oxide, nominally, UO2 or UO2±x. Titanate Systems In one embodiment, titanates may provide a host phase for uranium and actinides, fission products and impurities present in wastes, including potentially toxic metals such as lead. The bulk of the sludge contains Mg, typically present as Mg(OH)2, MgO, MgCO3 or Mg metal is reacted during the process=MgTiO3, Mg2TiO4, MgTi2O5 MgO+TiO2═MgTiO3 2MgO+TiO2═Mg2TiO4 MgO+2TiO2═MgTi2O5 For any Ca impurity in the sludge and Sr fission product radioisotope perovskite can form, nominally, CaTiO3—SrTiO3, [(Ca,Sr)TiO3]. This phase can also incorporate other fission products and impurities into its structure. Uranium that is present in the sludge in the sludge can react with titanium oxide and other components to form: Pyrochlore nominally A2B2O7-x, were A is nominally U, Th, actinides, Ca, rare earths, Y Ti3+; and B is nominally Ti, Al, Zr, Mg and transition metals. Cl can be accommodated on the O site. Traces of fission products may also be found in the structure. Brannerite nominally UTi2O6 but with Ca, rare earths, Zr, Y, Th and other fission products substituted for U; and Zr, Mg, Al, Ru, and transition metals such as Fe, Ni, Cr, Tc and other fission products substituted for Ti. Zirconolite nominally ACB2O7-x, were A is nominally Ca with U, Th, actinides, rare earths, Y as possible substitutions, C is nominally Zr with Hf, U, Th, rare earth, Ti3+; and B is nominally Ti, Al, Zr, Mg and transition metals such as Fe, Cr and Ni. Cl can be accommodated on the O site. Traces of fission products may also be found in the structure. Titanium dioxide is also present as a catalyst for assisting in the decomposition of salts, nitrates, organics and hydrocarbons during processing, in particular during calcination. Silicate In one embodiment, the addition of silica as silicon oxide (quartz, cristobalite), a silica sol, a glass frit, or a zeolite or other silicate mineral to form a wasteform substantially composed of magnesium silicate and silicate glass is another route that has been shown to produce dense durable wasteforms. The target phase is forsterite (Mg2SiO4) but the wasteform system can also include and MgSiO3 (enstatite) and CaMgSi2O6 phases. Some Mg can also be incorporated into the silicate glass phase, which bonds the mineral phases. In these systems MgO can also be present, which is encapsulated by the mineral silicates and glass phases. To obtain a high waste loading, wastes have been processed with additives in a way such that a coarse fraction of MgO remains. The coarse fraction is encapsulated in the glass-mineral phases in the wasteform. This is achieved by firstly restricting the amount of particle size reduction that occurs in the front-end mixing stage, calcining the waste plus additives to form MgO grains and then limiting the consolidation temperature to 900-1050° C. to retain the desired mineralogy and morphology in the wasteform. Aluminate In one embodiment, Mg is accommodated in a spinel (MgAl2O4), some impurities such as Fe, Cr Ni, etc. may also be present in the spinel structure. In this system silica or silicate additions are added to form a glass bonding phase that incorporates the fission products and other ions in the waste sludge. Multicomponent Systems In various embodiments, multicomponent systems may be used. Non-limiting examples of such systems include: titanate+silicate+aluminate; titanate+silicate; silicate+aluminate; titanate+aluminate phase combinations have been shown to be more flexible. In one embodiment, dense, durable titanate plus silicate systems bonded with a silicate glass have been made and these form stable, durable wasteforms. Additional Additives The use of a number of additives as redox control agents and mineralizers have been tested. These include the use of titanium, nickel and inconel powders to maintain reducing conditions during HIPing and act as a potential sink for volatile species during reaction. Wollastonite and cryolite have also been used as mineralizers to aid in the formation of glass ceramics. These additives have been tested individually and in combination at up to 10 wt % in the compositions outlined in Table 2. TABLE 2ReagentPlannedMagnoxWaste Simulant(wt %-(P)sludgeClinoptilolite/sandGlassSilicaInactivecalcined) CPSSimulant(9:1)TiO2fritsandClino#160—3010——#260151510——#360—1525——#46030—10——#56015—25——#660——40——#7—60201010——#8—404020———#9——100—#1045—4510#1160—3010#1235203510#13P40TBCTBCTBCTBC—#14P80TBCTBCTBCTBC—#16P100—————#17P—100————#185030—20——#196020—20——#20602020———#216020—12—8#22CPS60151510——#23CPS40—4020——#24CPS40———60—#25CPS57———43—#26CPS66———33—#27CPS40—66———#28CPS57—50———#29CPS66—40———#30CPS4075————#31CPS5750————#32CPS6625————#33CPS502525———#34CPS602020———##35CPS701515——— In various embodiment, additional densification aids such as lithium salts, lithium silicate, sodium salts, sodium silicates and fluorides (such as sodium fluorosilicate flux), and calcium fluoride added as individual components or as part of a frit, may be used, as they may aid in densification. Other embodiments of the invention will be apparent to those skilled in the art from consideration of the specification and practice of the invention disclosed herein. It is intended that the specification and examples be considered as exemplary only, with the true scope of the invention being indicated by the following claims. |
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053612850 | abstract | A tool system features an EDM crack removal head which can be positioned within a nozzle/thermal sleeve and then oscillated in a vertical and horizontal directions to enable precise material removal, leaving the surface free of cracks and in a condition for rewelding if required. If a crack is found in a surface which is covered by a thermal sleeve, then a small specialized EDM cutting head, which can be delivered from the under side of the reactor head and has an ability to pass through the ID of the thermal sleeve, is used. The head enables the sleeve to be cut and allow a portion of the sleeve to be removed. The crack can then be removed using a crack removal type of head. |
claims | 1. A method for generating a scatter image of an object at a projection angle in an imaging system, the method comprising:a. acquiring a non-grid image of the object using a radiation source and a detector;b. positioning a scatter rejecting aperture plate between the object and the detector at a first position, wherein the scatter rejecting aperture plate comprises a plurality of apertures, said plurality of apertures being positioned on a grid;c. acquiring a first grid image of the object with the scatter rejecting aperture plate disposed between the object and the detector at the first position;d. moving the scatter rejecting aperture plate to a second position between the object and the detector;e. acquiring a second grid image of the object with the scatter rejecting aperture plate disposed between the object and the detector at the second position;f. generating a scatter image of the object based on the non-grid image, the first grid image, and the second grid image; andg. storing the scatter image of the object. 2. The method of claim 1, further comprising moving the scatter rejecting aperture plate to at least a third position between the object and the detector; acquiring at least a third grid image of the object with the scatter rejecting aperture plate disposed between the object and the detector at the third position; and generating the scatter image of the object based on the non-grid image, the first grid image, the second grid image, and the at least third grid image. 3. The method of claim 1, wherein moving the scatter rejecting aperture plate comprises moving the scatter rejecting aperture plate uni-directionally. 4. The method of claim 3, wherein moving the scatter rejecting aperture plate uni-directionally comprises moving the scatter rejecting aperture plate at least one of horizontally and vertically. 5. The method of claim 1, wherein moving the scatter rejecting aperture plate comprises moving the scatter rejecting aperture plate bi-directionally. 6. The method of claim 5, wherein moving the scatter rejecting aperture plate bi-directionally comprises moving the scatter rejecting aperture plate horizontally and vertically. 7. The method of claim 1, wherein moving the scatter rejecting aperture plate comprises rotating the scatter rejecting aperture plate. 8. The method of claim 1, wherein a three-dimensional image of the object is generated based on the scatter image and wherein resolution of the three-dimensional image of the object is dependent on a plurality of scatter rejecting aperture plate positions. 9. The method of claim 8, wherein increasing the number of scatter rejecting aperture plate positions increases the resolution of the three-dimensional image of the object. 10. The method of claim 1, further comprising automatically moving the scatter rejecting aperture plate between positions. 11. The method of claim 1, wherein the scatter rejecting aperture plate has a shape selected from the group comprising hexagonal, rectangular, and circular. 12. A method for generating a three-dimensional image of an object, the method comprising:a. acquiring a plurality of projection images of the object using a source and a detector oriented at a plurality of projection angles relative to the object; said plurality of projection angles being realized by relatively rotating the object and the radiation source in a common plane of rotation;b. acquiring a scatter image at each of the plurality of projection angles, wherein acquiring each scatter image comprisesi. acquiring a non-grid image of the object using a radiation source and a detector,ii. positioning a scatter rejecting aperture plate between the object and the detector at a first position, wherein the scatter rejecting aperture plate comprises a plurality of apertures, said plurality of apertures being positioned on a grid,iii. acquiring a first grid image of the object with the scatter rejecting aperture plate disposed between the object and the detector at the first position,iv. repositioning the scatter rejecting aperture plate to a second position between the object and the detector,v. acquiring a second grid image of the object with the scatter rejecting aperture plate disposed between the object and the detector at the second position, andvi. generating a scatter image of the object based on the non-grid image, the first grid image, and the second grid image;c. generating a plurality of scatter free projection images by correcting the plurality of projection images based on respective ones of a plurality of stored scatter images by subtracting the scatter images from the respective projection images in a single process step; andd. reconstructing a three-dimensional image of the object based on the scatter free projection images. 13. The method of claim 12, wherein repositioning the scatter rejecting aperture plate comprises moving the scatter rejecting aperture plate from a first position to at least a second position between the object and the detector. 14. The method of claim 13, wherein moving the scatter rejecting aperture plate comprises one selected from the group comprising moving the scatter rejecting aperture plate uni-directionally, moving the scatter rejecting aperture plate bi-directionally, and rotating the scatter rejecting aperture plate. 15. The method of claim 12, wherein repositioning the scatter rejecting aperture plate comprises moving the sample from a first position to at least a second position in front of the scatter rejecting aperture plate. 16. The method of claim 15, wherein moving the sample comprises one selected from the group comprising moving the sample uni-directionally, moving the sample bi-directionally, and rotating the sample. 17. A volumetric CT system for imaging an object configured to generate a scatter free image of an object for use in generating a three-dimensional image of the object, the system comprising:a. a source and a detector configured to move with respect to the object, the detector configured to acquire a plurality of images of the object;b. an aperture plate configured to be positioned at a plurality of positions between the object and the detector, the aperture plate comprising a plurality of apertures, said apertures being positioned on a grid; andc. a processor configured to acquire a non-grid image of the object without the aperture plate and a grid image of the object with the aperture plate at each of the plurality of positions between the object and the detector and generate the scatter image of the object based on the non-grid images and the grid images acquired at each of the plurality of positions. 18. The system of claim 17, wherein resolution of the three-dimensional image of the object is dependent on a number of positions of the aperture plate at each projection angle and wherein the resolution of the three-dimensional image of the object increases as the number of positions of the aperture plate at each projection angle increases. 19. The system of claim 17, wherein repositioning the aperture plate comprises one selected from the group comprising moving the aperture plate from a first position to at least a second position between the object and the detector and moving the sample from a first position to at least a second position in front of the aperture plate. 20. The system of claim 19, wherein moving the aperture plate or the sample comprises one selected from the group comprising moving the aperture plate or the sample uni-directionally, moving the aperture plate or the sample bi-directionally, and rotating the aperture plate or the sample. |
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abstract | A beam filter positioning device includes a first and a second axes operable to move a body supporting one or more collimators, one or more photon flattening filters, one or more electron foils, and field light mirror etc. The collimators may be configured to collimate radiation to define a treatment beam suitable for radiosurgery. A controller is programmed to control the servo motor of the first and second axes to accurately position the beam filters. Radiation apparatuses and systems incorporating the beam filter positioning device or assembly are also provided. |
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abstract | An underground ventilated system for storing nuclear waste materials. The system includes a storage module having an outer shell defining an internal cavity and an inner shell. A majority of the height of the outer shell may be disposed below grade. The outer shell may include a hermetically sealed bottom. First and second canisters are positioned in lower and upper portions within the cavity respectively in vertically stacked relationship. A centering and spacing ring assembly is interspersed between the first and second canisters to transfer the weight of the upper second canister to the lower first canister. The assembly may include centering lugs which laterally restrain the first and second canisters in case of a seismic event. A natural convection driven ventilated air system cools the canisters to remove residual decay heat to the atmosphere. In one non-limiting embodiment, the shells are made of steel. |
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048270636 | description | Referring now to the figures of the drawings in detail and first, particularly, to FIG. 1 thereof, there is seen a nuclear reactor fuel assembly intended for a pressurized water reactor having two square retainer plates 2 and 3 made of metal, which are parallel to one another. FIG. 1 also shows two guide tubes 4 and 5 made of metal, each of which accommodates one control rod. The longitudinal axes of the guide tubes pass through the two retainer plates 2 and 3 at an angle of 90.degree. and are each screwed firmly to one of the retainer plates 2 and 3 at each end. Each guide tube 4 and 5 is guided through a square space or mesh opening in square, lattice-like spacers 6 having outer and inner sheet-metal ribs 6a and 6b, which are located along the length of the guide tube 4 between the two retainer plates 2 and 3 and are retained in a force-locking manner, such as by being firmly welded on the guide tubes 4 and 5. A force-locking connection is one which connects two elements together by force external to the elements, as opposed to a form-locking connection which is provided by the shapes of the elements themselves. Fuel rods 8 are guided through each of the other spaces or mesh openings of the spacers 6 and are parallel to the guide tubes 4 and 5. Each fuel rod 8 is substantially formed of a cladding tube filled with nuclear fuel and closed in a gas-tight manner at both ends. The fuel rods 8 are not secured to either of the two retainer plates 2 and 3. Instead, the fuel rods 8 are retained in the spaces or mesh openings of the spacers 6 in an elastic manner, that is in a force-locking manner, by means of non-illustrated springs and rigid bearing nubs of the spacers 6. The fuel rods 8 therefore have play in the direction of the longitudinal axes thereof between the two retainer plates 2 and 3 so that they can expand without hindrance in the longitudinal direction, that is in the longitudinal direction of the fuel assembly. The spacers 6, have no twisted turbulence-promoting vanes protruding beyond the sides of the spaces or mesh openings, particularly on the inner ribs 6b. The fuel assembly has an additional lattice 7 between each two spacers 6, each of which is unequally spaced apart from the two adjacent spacers 6. The additional lattices 7 have sheet-metal ribs 9 and 10 which pass through each other at right angles and on edge, forming square spaces or mesh openings, each of which accommodates one of the fuel rods 8 or a guide tube 4 or 5. While the fuel rods 8 are guided in the spaces or mesh openings of the additional lattice 7 in such a way as to be spaced apart from the sheet-metal ribs 9 and 10, that is loosely and with play, a metal sheath 11 is mounted on the guide tubes 4 and 5 by welding on the inside surface to the guide tube 4 or 5. The outer surface of the metal sheath 11 is in turn welded to four sheet-metal ribs 9 and 10 at a time, so as to form the space or mesh opening in the additional lattice 7 in which the guide tube 4 or 5 is located. As FIGS. 2 and 4 show particularly clearly, the ribs 9 that are parallel to one another have leading and trailing edges 9a and 9b, which have zig-zag portions 9c and 9d. The zig-zag portions 9c and 9d are compactly located in the plane of the associated sheet-metal rib 9 and thus of the sides of the spaces or mesh openings of the additional lattice 7 that are formed by the rib 9. The zig-zag portions 9c and 9d of the leading and trailing edges 9a and 9b of the ribs 9 are also parallel to one another. A zig-zag portion 9c having legs of equal length that form an angle of 90.degree. with one another, is located on the leading edge 9a between each two mutually parallel ribs 10. The zig-zag portions 9c are located in such a way as to face the oncoming flow direction for the coolant in a nuclear reactor. A zig-zag portion 9d which also has legs of equal length is disposed on the leading edge 9b of the ribs 9 at each of the ribs 10. One of the legs is disposed on each either side of the associated rib 10, forming an angle of 90.degree. between the two legs. Each of the zig-zag portions 9d is located in the outflow direction of the coolant flwwing through the fuel assembly in a nuclear reactor, so that the zig-zags 9c and 9d of the mutually parallel leading and trailing edges 9a and 9b of the mutually parallel sheet-metal ribs 9 of the additional lattice 7 are staggered with respect to one another. The mutually parallel ribs 10 of the additional lattice 7 have turbulence-promoting vanes 12a and 12b on mutually parallel trailing edges 10a thereof. Each two turbulence-promoting vanes 12a and 12b which are mounted next to one another face away from one another, as shown in particular in FIG. 4. This is due to the fact that they are twisted about the longitudinal direction of the fuel rods 8 and the guide tubes 4. Each edge of each space or mesh opening of the additional lattice 7 having one fuel rod 8, has two turbulence-promoting vanes 12a and 12b, each of which comes to a point at the end of the edge of the space or mesh opening and protrudes beyond the side of the space or mesh opening associated with this edge. Each side of a space or mesh opening is formed by one rib 10. As FIG. 4 shows, bending lines or deflection curves 12c and 12d at which the turbulence-promoting vanes 12a and 12b begin to protrude beyond the side of the space or mesh opening of the additional lattice 7, are of equal length and each forms the same angle .alpha. with the associated trailing edge 10a of the ribs 10. Both bending lines 12c and 12d, like the turbulence-promoting vanes 12a and 12b, are located on the outside above the trailing edge 10a of the ribs 10. Turbulence-promoting vanes of this kind may also be located on the mutually parallel leading edge of the ribs 10, which is parallel to the trailing edge 10a. The ribs 9 and 10 of the additional lattice 7 each have brackets 13 and 14 formed on the two ends thereof, which grip the additional lattice 7 inbetween the fuel rods 8 that are located at outer regions 15 and 16 of the fuel assembly. Furthermore, the sides of the ribs 9 and 10 that form the surface of the additional lattice 7 in the spaces and that are parallel to the longitudinal direction of the fuel rods 8 and the control rod guide tube 4, are advantageously smooth and/or flat (such as by dispensing with bearing nubs). This is done so that they optimally present little resistance to a flow of coolant in the longitudinal direction of the fuel assembly in a nuclear reactor. The foregoing is a description corresponding in substance to German Application No. P 36 32 627.5, dated Sept. 25, 1986, the International priority of which is being claimed for the instant application, and which is hereby made part of this application. Any material discrepancies between the foregoing specification and the aforementioned corresponding German application are to be resolved in favor of the latter. |
abstract | It is desirable to achieve a co-incident investigative kV source for a therapeutic MV source—a so-called “beams-eye-view” source. It has been suggested that bremsstrahlung radiation from an electron window be employed; we propose a practical structure for achieving this which can switch easily between a therapeutic beam and a beam-eye-view diagnostic beam capable of offering good image resolution. Such a radiation source comprises an electron gun, a pair of targets locatable in the path of a beam produced by the electron gun, one target of the pair being of a material with a lower atomic number than the other, and an electron absorber insertable into and withdrawable from the path of the beam. In a preferred form, the electron gun is within a vacuum chamber, and the pair of targets are located at a boundary of the vacuum chamber. The lower atomic number target can be Nickel and the higher atomic number target Copper and/or Tungsten. The electron absorber can be Carbon, and can be located within the primary collimator, or within one of a plurality of primary collimators interchangeably locatable in the path of the beam. Such a radiation source can be included within a radiotherapy apparatus, to which the present invention further relates. A flat panel imaging device for this source can be optimised for low energy x-rays rather than high energy; Caesium Iodide-based panels are therefore suitable. |
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056125437 | abstract | A basket for transporting, storing, and containing nuclear fuel assemblies having an assembly of sleeves with a plurality of sleeves arranged in a uniform pattern and secured within a cylindrical shell. Each of the plurality of independent sleeves being sized to secure and contain a fuel assembly. A plurality of alternating sleeves of the plurality of independent sleeves are configured to include an angular shaped separator element secured to each corner of each of the plurality of alternating sleeves. A sheet of neutron absorbing material is positioned between each of the plurality of alternating sleeves for maintaining fission reactions within the basket below a critical level necessary to sustain a fission reaction. A support element for positioning and securing the plurality of independent sleeves is secured within the cylindrical shell. A bottom plate is secured to the bottom of the cylindrical shell providing vertical support for the plurality of independent sleeves. A shield lid is secured to the cylindrical shell and includes a plurality of disc elements and an access port for selective entry into the basket and a lid element is secured to the shield lid and to the cylindrical shell. The lid element including an access port for selective entry into the basket. |
summary | ||
claims | 1. A passive hydrogen recombiner and igniter comprising:a substantially horizontal, flat metallic plate having an underside coated with a hydrogen recombination catalyst, supported in a peripheral housing having a first gaseous intake below the substantially horizontal, flat metallic plate and a first gaseous outlet around a periphery of the substantially horizontal, metallic plate;a second gaseous intake through the housing and through a first set of swirl vanes substantially proximate to and in communication with an upper side of the substantially horizontal, flat metallic plate, with the first set of swirl vanes configured to create a vortex out of a second gas traversing the second gaseous intake;a second gaseous outlet through an upper portion of the housing through which the vortex exits;a first passive igniter supported proximate the first gaseous intake; anda second passive igniter supported proximate the second gaseous outlet. 2. The passive hydrogen recombiner and igniter of claim 1 wherein the hydrogen recombination catalyst is either platinum or palladium or a combination thereof. 3. The passive hydrogen recombiner and igniter of claim 1 wherein the underside of the substantially horizontal, flat metallic plate includes downwardly projecting vanes covered with the hydrogen recombination catalyst, structured to direct a first gas entering the first gaseous intake to the first gaseous outlet. 4. The passive hydrogen recombiner and igniter of claim 1 wherein the first gaseous outlet extends up through an interior of at least some of the first set of swirl vanes and exits outside the second gaseous outlet. 5. The passive hydrogen recombiner and igniter of claim 4 wherein the first set of swirl vanes are structured to transfer heat from the first gas traveling through the swirl vanes to the second gas entering the second gaseous intake. 6. The passive hydrogen recombiner and igniter of claim 1 wherein the first igniter is a platinum or palladium wire. 7. The passive hydrogen recombiner and igniter of claim 6 wherein the wire is wound as a spring to increase its surface area. 8. The passive hydrogen recombiner and igniter of claim 1 wherein the second igniter is powered by the vortex. 9. The passive hydrogen recombiner and igniter of claim 8 wherein the second igniter is a rotating device that accumulates charge, similar to a van de Graf generator, to create a spark as an ignition activation energy. 10. The passive hydrogen recombiner and igniter of claim 8 wherein the second igniter is a rotating device that drives an electric generator that charges a capacitor, which is structured to throw a spark once a particular voltage is reached. 11. The passive hydrogen recombiner and igniter of claim 8 wherein the second igniter is a rotating device that drives a piezoelectric device to create a spark. 12. The passive hydrogen recombiner and igniter of claim 1 wherein the second gaseous outlet includes a cover spaced from the second gaseous outlet so the second gas can exhaust from under the cover. 13. The passive hydrogen recombiner and igniter of claim 1 wherein an upper side of the substantially horizontal, flat metallic plate has a second set of swirl vanes attached to its surface. 14. The passive hydrogen recombiner and igniter if claim 13 wherein the second set of vanes are co-directional with the vortex. 15. The passive hydrogen recombiner and igniter of claim 1 wherein an upper side of the substantially horizontal, flat metallic plate is substantially covered with a hydrogen recombination catalyst. |
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06278756& | abstract | The invention relates to a sensor for a measuring an electrochemical corrosion potential comprising a sensor tip, a conductor electrically connected to the sensor tip, an insulating member which surrounds the conductor, a connecting member which surrounds the conductor; and a sleeve which fits over the sensor tip, the insulating member, and the connecting member, the sleeve having inner threads which engage with corresponding outer threads on at least one of the sensor tip and the connecting member. |
abstract | A mirror that has a mirror substrate (12), a reflection layer stack (21) reflecting electromagnetic radiation incident on the optical effective surface (11), and at least one piezoelectric layer (16) arranged between the mirror substrate and the reflection layer stack and to which an electric field for producing a locally variable deformation is applied by way of a first electrode arrangement and a second electrode arrangement situated on alternate sides of the piezoelectric layer. In one aspect, both the first and the second electrode arrangements have a plurality of electrodes (20a, 20b), to each of which an electrical voltage relative to the respective other electrode arrangement can be applied via leads (19a, 19b). Separate mediator layers (17a, 17b) set continuous electrical potential profiles along the respective electrode arrangement, and where said mediator layers differ from one another in their average electrical resistance by a factor of at least 1.5. |
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041397770 | abstract | A cyclotron suitable for use in neutron therapy, and comprising a pair of opposed, spaced pole shoes having their adjacent inner surfaces defining an accelerator zone, an electromagnetic coil system, at least one hollow accelerating dee electrode positioned in the accelerator zone, and having a radio-frequency resonator associated therewith, a magnet yoke shaped to substantially enclose the accelerator zone and constitute a neutron attenuation shield for neutrons produced in the cyclotron, a vacuum chamber enclosing the accelerator zone and each dee electrode, means for providing charged particles for acceleration within the accelerator zone, a target zone for a target device, and a neutron beam outlet in the magnet yoke for emission of a neutron beam produced in the cyclotron. The cyclotron includes auxiliary neutron shield means in the forward peak zone of a neutron beam produced in the cyclotron to attenuate neutron and gamma radiation in the forward peak zone to a patient tolerable level. Each radio-frequency resonator is enclosed within the magnet yoke. The cyclotron may have pivot means for pivotally mounting the cyclotron to allow variation of the direction of a neutron beam during use. The disclosure further relates to a neutron therapy installation incorporating such a cyclotron. |
abstract | Methods implemented by at least one electronic processor for generating pointwise fast neutron spectra may include receiving composition data; receiving source data or calculating the source data; receiving nuclear data; and calculating the pointwise fast neutron spectrum based on numerical integration using the composition, source, and nuclear data. Systems for generating pointwise fast neutron spectra may include a bus; at least one electronic processor connected to the bus; an input device connected to the bus; and a communication link connected to the bus. The at least one electronic processor may be configured to receive composition data from the input device via the bus, to receive source data from the input device via the bus or to calculate the source data, to receive nuclear data from the communication link via the bus, and to calculate the pointwise fast neutron spectrum based on numerical integration using the composition, source, and nuclear data. |
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description | FIG. 1 is selective view of a related art nuclear core shroud 10, useable in a nuclear reactor like a BWR. Core shroud 10 may be positioned in annulus region 20, which is an annular space formed between shroud 10 and an inner wall of a reactor pressure vessel (not shown) that receives fluid coolant flow and directs it downward for entry at a bottom of core 30. Shroud 10 may be a cylindrical structure surrounding core 30 that partitions the reactor into these downward and upward coolant flows on opposite radial sides of shroud 10. One or more jet pump assemblies 40 may line annulus 20 and direct coolant flow in this manner. After being directed downward past core shroud 10, coolant may then flow up through core 30 inside shroud 10. Core 30 is typically populated by several fuel assemblies (not shown) generating heat through nuclear fission during operation, and the coolant exiting core 30 may be quite energetic and potentially boiling. This energetic fluid flows up through and out of core 30 and shroud 10, potentially into steam separating and drying structures and ultimately to a turbine that drives a generator to convert the energetic flow into electricity. The top portion 15 of shroud 10 may terminate below such drying structures, and various core equipment may rest on or join to shroud 10 about top portion 15, which may be called a steam dam. One or more gussets 16 may be aligned about top portion 15 of shroud 10 to support or join a shroud head (not shown), chimney, or drying structures. During a reactor outage, such as a refueling outage or other maintenance period, the reactor vessel may be opened and inspected, and internal structures of vessel may be removed. During an outage, loading equipment such as a bridge and trolley above the reactor, and 40-50 feet above core 30 and shroud 10, may move and load new fuel assemblies in core 30. Visual inspections of shroud 10, core 30, and/or any other component can be accomplished with video or camera equipment operated from the bridge or other loading equipment above the reactor during this time. For example, the positioning and inspection devices of co-owned US Pat Pub 2017/0140844 to Kelemen, published May 18, 2017, incorporated herein by reference in its entirety, may be used in connection with inspections from steam dam 15. Example embodiments include apparatuses, devices, and systems including the same for moving inspection and tooling articles in congested underwater areas, such as nuclear reactors undergoing maintenance, refueling, or inspection. Example embodiments are operable from a far edge of the reactor, such as on a steam dam or outer wall of the reactor. An apparatus shoulder can mount at this edge, such as via a clamping device secured to a steam dam. The shoulder may be rotatable and include arms that are further rotatable. A tool, reactor component, and/or inspection device like a camera or VARD can be secured at the end of the arm opposite the shoulder. This rotatable extension may permit a relatively lightweight device to extend several meters transversely into a reactor without requiring any interaction from refueling apparatuses. In light of this extension, a float may be positioned away from the shoulder so as to counter torque on the shoulder from the arms and article(s) when immersed in water. Rotation of the shoulder and arms may be via vertical axes, so that the apparatus remains in a single axial plane or height in the reactor. The rotation may be accomplished with manual rotation, such as with a handling pole, or with local motors in example embodiment apparatuses. The arms may also rotate so as to overlap, permitting compact installation and removal. Power, controls, and/or data may be provided underwater to example embodiment apparatuses through an umbilical connection to operators. Because this is a patent document, general, broad rules of construction should be applied when reading it. Everything described and shown in this document is an example of subject matter falling within the scope of the claims, appended below. Any specific structural and functional details disclosed herein are merely for purposes of describing how to make and use examples. Several different embodiments and methods not specifically disclosed herein may fall within the claim scope; as such, the claims may be embodied in many alternate forms and should not be construed as limited to only examples set forth herein. It will be understood that, although the ordinal terms “first,” “second,” etc. may be used herein to describe various elements, these elements should not be limited to any order by these terms. These terms are used only to distinguish one element from another; where there are “second” or higher ordinals, there merely must be that many number of elements, without necessarily any difference or other relationship. For example, a first element could be termed a second element, and, similarly, a second element could be termed a first element, without departing from the scope of example embodiments or methods. As used herein, the terms “and,” “or,” and “and/or” include all combinations of one or more of the associated listed items unless it is clearly indicated that only a single item, subgroup of items, or all items are present. The use of “etc.” is defined as “et cetera” and indicates the inclusion of all other elements belonging to the same group of the preceding items, in any “and/or” combination(s). It will be understood that when an element is referred to as being “connected,” “coupled,” “mated,” “attached,” “fixed,” etc. to another element, it can be directly connected to the other element, or intervening elements may be present. In contrast, when an element is referred to as being “directly connected,” “directly coupled,” etc. to another element, there are no intervening elements present. Other words used to describe the relationship between elements should be interpreted in a like fashion (e.g., “between” versus “directly between,” “adjacent” versus “directly adjacent,” etc.). Similarly, a term such as “communicatively connected” includes all variations of information exchange and routing between two electronic devices, including intermediary devices, networks, etc., connected wirelessly or not. As used herein, the singular forms “a,” “an,” and “the” are intended to include both the singular and plural forms, unless the language explicitly indicates otherwise. Indefinite articles like “a” and “an” introduce or refer to any modified term, both previously-introduced and not, while definite articles like “the” refer to a same previously-introduced term; as such, it is understood that “a” or “an” modify items that are permitted to be previously-introduced or new, while definite articles modify an item that is the same as immediately previously presented. It will be further understood that the terms “comprises,” “comprising,” “includes,” and/or “including,” when used herein, specify the presence of stated features, characteristics, steps, operations, elements, and/or components, but do not themselves preclude the presence or addition of one or more other features, characteristics, steps, operations, elements, components, and/or groups thereof. The structures and operations discussed below may occur out of the order described and/or noted in the figures. For example, two operations and/or figures shown in succession may in fact be executed concurrently or may sometimes be executed in the reverse order, depending upon the functionality/acts involved. Similarly, individual operations within example methods described below may be executed repetitively, individually or sequentially, to provide looping or other series of operations aside from single operations described below. It should be presumed that any embodiment or method having features and functionality described below, in any workable combination, falls within the scope of example embodiments. As used herein, “axial” and “vertical” directions are the same up or down directions oriented along the major axis of a nuclear reactor, often in a direction oriented with gravity. “Transverse” directions are perpendicular to the “axial” and are side-to-side directions oriented in a single plane at a particular axial height. The Inventors have newly recognized a need for reliable and non-interfering positioning in remote operations, including inspections. In some nuclear reactors during maintenance periods, cumbersome inspection equipment is operated from a refueling bridge or overhead crane and can interfere with refueling operations due to co-location with the refueling equipment. Furthermore, positioning with the fuel mast prevents inspection of other areas of a reactor not being refueled. While remotely-operated vehicles may avoid the problem of co-location with the refueling equipment, remotely-operated vehicles submerged in coolant do not have reliable positioning to ensure an inspection is at a verified position. Example embodiments described below uniquely enable solutions to these and other problems discovered by the Inventors. The present invention is systems and apparatuses for positioning articles without interfering with refueling operations in a nuclear reactor. In contrast to the present invention, the few example embodiments and example methods discussed below illustrate just a subset of the variety of different configurations that can be used as and/or in connection with the present invention. FIG. 2 is an illustration of an example embodiment reactor positioning and inspection apparatus 100. As seen in FIG. 2, apparatus 100 may work with positioning device 50 from the incorporated '844 application. That is, positioning device 50 may be secured to, and move along a perimeter of, a steam dam or other ledged top of a nuclear reactor, and example embodiment apparatus 100 may be mounted to device 50 along the same. For example, device 50 may include central hex receiver 51, and example embodiment apparatus 100 may include hex mounting post 105 that seats into receiver 51, or device 50 and apparatus 100 may be joined in any other fashion. In this way, apparatus 100 can be supported, secured, and movable about a nuclear reactor, fuel pool, or other area for inspection and other operations, such as during a maintenance or refueling period, including periods of submersion and open air operations, depending on flooding levels. Example embodiment positioning an inspection apparatus 100 includes shoulder 106, which may contain a large amount of its mass and be directly supported about a reactor or other inspection area. Shoulder 106 may include several powering and control features for other remote operations of apparatus 100, including motors 101 and 102 and associated transmissions, umbilical connection 107, mounting shaft 105, and installation guide 108. Umbilical connection 107 may connect data, power, pneumatics, controls, instrumentation, and/or any other external source to apparatus 100. For example, an electrical power and control signal line may be run through umbilical connection 107 underwater from operators above the reactor in open air on a refueling platform or bridge. Example embodiment positioning and inspection apparatus 100 includes two jointed arms—back arm 110 and articulating front arm 120. As shown in FIG. 2, back arm 110 may be secured to shoulder 106 and extend in a transverse direction out to elbow 115. Back arm 110 may be hollow so as to convey power and data from shoulder 106 to elbow 115 and beyond. Float 109 may be secured to back arm 110 toward elbow 115 to impart buoyancy and remove torque on shoulder 106 from weight of back arm 110. Float 109 provides a desired level of buoyancy in a fluid. As such, when immersed in a reactor coolant such as water, back arm 110 may carry power and instrumentation inside arm 110 and be relatively balanced and in an axial direction to avoid overly torqueing or rotating shoulder 106 and thus device 50 to which it may seat. Front arm 120 extends beyond back arm 110 and articulates with respect to back arm 110 at elbow 115. For example, elbow 115 may be a stacked hinge or joint that allows front arm 110 to rotate about a vertical axis through elbow 115. Elbow 115 may also transmit power, data, instrumentation, etc. from back arm 110 to front arm 120, and front arm 120 may be similarly hollow to carry the same, ultimately to hand 130. Both front arm 120 and back arm 110 may have square cross sections to enhance shearing resistance to loads carried at their ends. Front arm 120 terminates at hand 130 useable to carry and operate articles of interest such as tooling, instrumentation, components, etc. In the example of FIG. 2, a VARD 200 is clamped in hand 130 for video inspection of various core locations. Hand 130 may be capable of selective grasping via pneumatic cylinder 131 that opens and retracts hand 130. Pneumatic cylinder may be operated by the same instrumentation lines that rundown front arm 120 and back arm 110 from umbilical connection 107. Hand 130 may take on any desired shape or configuration to attach to or operate desired articles, and other operative devices besides pneumatic cylinder 131 may open, close, and otherwise operate hand 130, even when submerged in coolant. Camera mount 132 may separately hold another article, such as a camera working in association with VARD 200. FIG. 3 is an overview schematic view of example embodiment positioning and inspection apparatus 100, illustrating operation of the same. Operations may be performed while apparatus 100 is mounted to device 50 underwater at an edge of a flooded reactor. As shown in FIG. 3, apparatus 100 is rotatable at shoulder 106 in direction 126, which may be rotation about a vertical axis at shoulder 106. Similarly, apparatus 100 is rotatable at elbow 115 in direction 125, which may be rotation about a vertical axis at elbow 115. Through circumferential movement of device 50 in direction 127, apparatus 100 may reach across to any reactor position. As such, any tool held in hand 130 may be positioned at nearly any desired transverse or angular position in a reactor through movement of apparatus 100 in directions 125, 126, and/or 127. Moreover, through known degrees of rotation in directions 125 and 126, position in direction 127, and known lengths of arms 110 and 120, the transverse position of any article held in hand 130 may be known with extreme precision in a reactor. At maximum extension, with arms 110 and 120 straight, apparatus may reach a distance d, a radius that may potentially span several meters, such as 3 meters, nearing a radius of a commercial nuclear reactor. In this way instrumentation or inspection devices in hand 130 may be placed vertically overlapping any fuel position in a reactor. Similarly, apparatus 100 may collapse by rotating front arm 120 back through direction 125 to be partially or fully overlapping with back arm 110, presenting a much smaller transverse length of half d. Such a collapsed position may be useful during installation or removal, or during movement of device 50, to require less space. Apparatus 100 may also form any angle between front arm 120 and back arm 110, potentially reaching around other core structures or maintenance activities through rotation at elbow 115 and/or shoulder 106. Rotation in directions 126 and 125 may be accomplished under local force generated by apparatus 100 or through remote powering. For example, as shown in FIG. 2, shoulder motor 101 may be connected to mounting post 105 through a transmission such as a gear train in shoulder 106. Shoulder motor 101 may rotate shoulder 106, and thus the remainder of apparatus 100, about post 105, which may be pivotable in direction 126. Similarly, elbow motor 102 may be connected to a post in elbow 115 through a transmission such as a transmission belt extending through shoulder 106 and back arm 110. Elbow motor 102 may rotate front arm 120, and thus any device held in hand 130, in direction 125. Instead of using motors 101 and/or 102, rotation in directions 126 and 125 may be achieved through manual drives 111 and 112 (FIG. 2). For example, a handling pole or other actuator may mate with shoulder manual drive 111 and rotate the same to reproduce actuation of shoulder motor 101 and rotation about mast 105 in direction 126. Similarly, the handling pole or other actuator may mate with elbow manual drive 112 and rotate the same to reproduce actuation of elbow motor 102 and rotation about elbow 115 in direction 125. In this way, example embodiment apparatus 100 may retain mobility in directions 126 and 125 even through manual actuation. By being positioned at shoulder 106 and transferring power through transmissions, motors 101 and 102 may keep weight and operations at shoulder 106, near umbilical connection 107 that may control motors 101 and 102. Other commands and power may also be provided through umbilical connection 107 through shoulder 106, presenting a relatively compact control interface and most massive aspect overlapping with device 50 at a periphery of a nuclear reactor, such as at steam dam 15. For example, actuation of pneumatic cylinder 131, and/or actuation of motors 101 and 102 may all be instructed by operators through umbilical connection 107, to position and otherwise operate apparatus 100 in a desired manner. Alternatively or additionally, command, control, and power may be locally present in shoulder 106 and other components through the use of batteries and wireless communications. In these ways, operators potentially quite remote from apparatus 100, such as above a flooded level in open air, may directly control and receive data from apparatus 100 underwater. Example embodiment positioning and inspection apparatus 100 is fabricated of materials that are compatible with a nuclear reactor environment, including materials that maintain their physical characteristics when exposed to high-temperature fluids and radiation. For example, metals such as stainless steels and iron alloys, aluminum alloys, zirconium alloys, etc. are useable in shoulder 106, back arm 110, front arm 120, etc. Similarly, direct connections between components may be lubricated and fabricated of alternating or otherwise compatible materials to prevent seizing, fouling, or metal-on-metal reactions. By use of lighter-weight materials and hollow profiles, an example embodiment apparatus having a 1.5-meter back arm 100 may weigh 50 pounds or less. FIG. 4 is a profile schematic view of example embodiment positioning and inspection apparatus 100, illustrating an installation and/or removal of the same from an underwater environment. Although front arm 120 is shown as not fully withdrawn in FIG. 4, it is understood that front arm may be completely under back arm 110 and reaching back to shoulder 106 during installation or removal. As seen in FIG. 4, handling pole 201 may connect to install blade 108 through a lock-and-hook mechanism, auger-and-tang, mechanism, or any other selective fastener. When locked with install blade 108, handling pole 201 moves apparatus 100 with it, including in vertical direction 202. Handling pole 201 may be under control of an operator above the water level, either manually or through a crane or other powered connection. In combination with float 109, apparatus may thus be positioned at a desired vertical and transverse position through minimal effort via pole 201. For example, apparatus 100 may be joined to device 50 (FIG. 1) through installation actions and movement with handling pole 201 such that mast 105 seats into device 50. Of course, other installations and positionings of example embodiment apparatus 100 are easily achieved through proper motion of pole 201. Once installed, pole 201 may disengage from install blade 108 by releasing or unfastening from the same. For de-installation, handling pole 201 may be reattached to install blade 108 and removed from the environment in vertical direction 202. Example embodiments and methods thus being described, it will be appreciated by one skilled in the art that example embodiments may be varied and substituted through routine experimentation while still falling within the scope of the following claims. For example, any number of different bases and tools can be used in example embodiment positioning apparatuses, simply through proper dimensioning and shaping. Such variations are not to be regarded as departure from the scope of these claims. |
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description | This present application is a divisional application of co-pending U.S. application Ser. No. 16/662,523, filed 24 Oct. 2019, which claims priority to provisional application 62/749,875, filed Oct. 24, 2018, and which is a continuation in part (CIP) of U.S. application Ser. No. 15/488,983, filed Apr. 17, 2017, which claimed priority to U.S. application Ser. No. 14/190,389, filed Feb. 26, 2014, which has issued as U.S. Pat. No. 9,636,524 on May 2, 2017, which claimed priority to U.S. application Ser. No. 13/532,447, filed on Jun. 25, 2012, now abandoned, which claimed priority to provisional U.S. patent application 61/571,406 filed Jun. 27, 2011. This invention is in the technical area of apparatus and methods for Boron Neutron capture therapy for cancer. Boron Neutron Capture Therapy (BNCT) is not new in the art, as thermal neutrons have been used for cancer therapy for the destruction of cancer tumors. These neutrons interact with boron-10 that has been placed at the cancer site. The neutrons interact with the boron to produce fission events whereby alpha particles and lithium nuclei are created. These massive ionized particles are then released, destroying the chemical bonds of nearby cancer tumor cells. At present the neutrons created in a reactor or accelerator pass through a moderator, which shapes the neutron energy spectrum suitable for BNCT treatment. While passing through the moderator and then the tissue of the patient, the neutrons are slowed by collisions and become low energy thermal neutrons. The thermal neutrons undergo reactions with the boron-10 nuclei at a cancer site, forming compound nuclei (excited boron-11), which then promptly disintegrate to lithium-7 and an alpha particle. Both the alpha particle and the lithium ion produce closely spaced ionizations in the immediate vicinity of the reaction, with a range of approximately 5-9 micrometers, or roughly the thickness of one cell diameter. The release of this energy destroys surrounding cancer cells. This technique is advantageous since the radiation damage occurs over a short range and thus normal tissues can be spared. Gadolinium can also be considered as a capture agent in neutron capture therapy (NCT) because of its very high neutron capture cross section. A number of gadolinium compounds have been used routinely as contrast agents for imaging brain tumors. The tumors have absorbed a large fraction of the gadolinium, making gadolinium an excellent capture agent for NCT. Therefore, GNTC may also be considered as a variation in embodiments of the present invention. The following definitions of neutron energy ranges, E, are used frequently by those skilled in the art of producing and using neutrons for medical, commercial and scientific applications: Fast (E>1 MeV), Epithermal (0.5 eV<E<1 Mev) and Thermal (E<0.5 eV) neutrons. BNCT has the potential to treat previously untreatable cancers such as glioblastoma multiforme (GBM). In the US brain tumors are the second most frequent cause of cancer-related deaths for males under 29 and females under 20. GBM is nearly always fatal and has, until now, no known effective treatment. There are approximately 13,000 deaths per year due to primary brain tumors. If conventional medicine is used where the glioblast is excised, new tumors almost invariably recur, frequently far from the original tumor site. Effective radiation therapy, therefore, must encompass a large volume and the radiation must be uniformly distributed. Conventional radiation treatment is usually too toxic to be of use against GBM. For distributed tumors, effective radiation therapy must encompass a larger volume and the radiation must be uniformly distributed. This is also true of liver cancers. The liver is the most common target of metastases from many primary tumors. Primary and metastatic liver cancers are usually fatal, especially after resection of multiple individual tumors. The response rate for nonresectable hepatocellular carcinoma to traditional radiation treatment or chemotherapy is also very poor. However, recent results indicate that the thermal neutron irradiation of the whole liver with a 10B compound, to be bombarded with low-energy neutrons, could be a way to destroy all the liver metastases. Recent research in BNCT has shown that neutron capture therapy can be used to treat a large number of different cancers. BNCT has been found to be effective and safe in the treatment of inoperable, locally advanced head and neck carcinomas that recur at sites that were previously irradiated with traditional gamma radiation. Thus, BNCT could be considered for a wider range of cancers. BNCT holds such promise because the dose to the cancer site can be greatly enhanced over that produced by y-radiation sources. This is a consequence of the fact that the neutron-boron reaction produces the emission of short-range (5-9 um distance) radiation, and consequently normal tissues can be spared. In addition, boron can achieve a high tumor-to-brain concentration ratio, as much as ten or more, thereby preferentially destroying abnormal tissue. BNCT has been tested using either nuclear reactors or accelerators to produce the neutrons, which are not practical or affordable for most clinical settings. Reactors also do not produce an ideal neutron spectrum and are contaminated with y-radiation. Fusion generators produce fast neutrons from the deuterium-deuterium (DD) or the deuterium-tritium (DT) reactions and are, in general, smaller and less expensive than accelerators and reactors. Fast neutrons thus produced must be moderated or slowed down to thermal or epithermal neutron energies using, for example, water or other hydrogen bearing materials. The fusion neutron generator has three basic components: an ion source, an electron shield and an acceleration structure with a target. The ions are accelerated from the ion source to usually a titanium target using a high voltage potential of between 40 kV to 200 kV, which can be easily delivered by a modern high voltage power supply. An electron shield is usually disposed between the ion source and the titanium target. This shield is voltage biased to repel electrons being generated when the positive D+ ions that strike the titanium target. This prevents these electrons from striking the ion source and damaging it due to electron heating. The target uses a deuterium D+ or tritium T+ absorbing material such as titanium, which readily absorbs the D+ or T+ ions, forming a titanium hydride. Succeeding D+ or T+ ions strike these embedded ions and fuse, resulting in DD, DT or TT reactions and releasing fast neutrons. Prior attempts at proposing fusion generators required the use of the DT reaction with the need for radioactive tritium and high acceleration powers. High yields of fast neutrons/sec were needed to achieve enough thermal neutrons for therapy in a reasonable length of time of therapy treatments. These prior schemes for achieving epithermal neutron fluxes are serial or planar in design: a single fast neutron generator is followed by a moderator, which is followed by the patient. Unfortunately, since the neutrons are entering from one side of the head, the planar neutron irradiation system leads to a high surface or skin dosage and a decreasing neutron dose deeper into the brain. The brain is not irradiated uniformly, and cancer sites have lower thermal neutron dosage the further they are from the planar port. A conventional planar neutron irradiation system 14 and its operation is shown in FIG. 1 labeled Prior Art. Conversion of fast neutrons 22 to thermal neutrons 30 takes place in a series of steps. First the fast neutrons 22 are produced by a cylindrical fast neutron generator 20 and then enter a moderating means 18 where they suffer elastic scatterings (collisions with nuclei of the moderating material's atoms). This lowers the fast neutrons to epithermal neutron 24 energies. A mixture of epithermals 24 and thermal neutrons 30 are emitted out of a planar port 16 and then enter the patient's head 26. The epithermal neutrons 24 are moderated still further in the patient's brain and moderated further to thermal neutrons, finally being captured by the boron at the tumor site. The fission reaction occurs, and alpha and Li-7 ions are released, destroying the tumor cells. The epithermal and thermal neutrons reach the patient's head through a planar port 16 formed from neutron absorbing materials that form a collimating means 28. The thermal and epithermal neutrons strike the patient's head on one side, and many neutrons escape or are not used. One escaping neutron 38 is shown as representative. This is an inefficient process requiring a large number of fast neutrons to be produced in order to produce enough thermal neutrons for reasonable therapy or treatment times (e.g. 30 min). To achieve higher yields of fast neutrons the planar neutron irradiation system 14 requires that one use either the DD fusion reaction with extremely high acceleration powers (e.g. 0.5 to 1.5 Megawatts) or the DT reaction which has an approximate 100-fold increase in neutron yield for the same acceleration power. The use of tritium has a whole host of safety and maintenance problems. Tritium gas is radioactive and extremely difficult to eliminate once it gets on to a surface. In the art of producing fast neutrons this requires that the generator be sealed and have a means for achieving a vacuum that is completely sealed. The generator head cannot be easily maintained and usually its lifetime is limited to less than 2000 hours. This reduces the possible use of this generator for clinical operation since the number of patients who could be treated would be small before the generator head would need replacement. On the other hand, the use of the DD fusion reaction allows one skilled in the art to use an actively-pumped-vacuum means with roughing and turbo pumps. The generator can then be opened for repairs and its lifetime extended. This makes the DD fusion reaction neutron generator optimum for clinical use. The downside for the DD fusion reaction is that high acceleration powers are required to achieve the desired neutron yield required by prior art methods. Improving the efficiency of producing the right thermal neutron flux at the cancer site is imperative for achieving BNCT in a clinical and hospital setting. In one embodiment a treatment system for evaluating boron sources for boron neutron cancer therapy (BNCT) s provided, comprising a substantially square secondary moderator having a central treatment chamber for a subject; and four substantially identical neutron generators, each comprising a pre-moderator block of moderating material having an upper surface, a lower surface, a first and a second end, opposite parallel side surfaces, a first length, a first width substantially less than the first length, and a first thickness, a cylindrical acceleration chamber having a first diameter substantially the first width of the pre-moderator block, sealed at one end to the upper surface of the pre-moderator block adjacent the first end of the pre-moderator block, with a vertical axis perpendicular to the upper surface, the acceleration chamber having a height and a top cover at a second end away from the pre-moderator block, a vacuum pump engaging the acceleration chamber at a right angle to the vertical axis, evacuating the acceleration chamber to a moderately high vacuum, a plasma ion chamber opening into the acceleration chamber through an ion extraction iris through the top cover of the acceleration chamber on the vertical axis of the acceleration chamber, a gas source providing deuterium gas to the plasma ion chamber, a microwave energy source ionizing the gas in the plasma ion chamber, a cylindrical primary isolation well extending a substantial distance into the pre-moderator block from the upper surface, centered on the vertical axis of the acceleration chamber, a secondary isolation well substantially in a shape of a hollow cylinder surrounding the primary isolation well, to a depth somewhat less than the substantial distance of the primary isolation well, within the first diameter of the acceleration chamber, a water-cooled titanium target disk having a target surface orthogonal to the axis of the acceleration chamber, the target disk having a diameter substantially smaller then a diameter of the isolation well, positioned at a lower extremity of the isolation well, the target disk biased to a substantial negative DC voltage, and electrically grounded metal cladding covering all otherwise exposed surfaces of the pre-moderator block. The four neutron generators are positioned around the substantially square secondary moderator with the axis of each acceleration chamber passing through the center of the treatment chamber. In one embodiment the secondary moderator is a block or blocks of solid moderator material. Also, in one embodiment the secondary moderator is a container filled with heavy water. And in one embodiment the secondary moderator is a container filled with granulated moderator material. In the following descriptions reference is made to the accompanying drawings that form a part hereof, and in which are shown by way of illustration specific embodiments in which the invention may be practiced. It is to be understood that other embodiments may be utilized, and structural changes may be made without departing from the scope of the present invention. Uniform Delivery of Thermal Neutrons to the Cancer Sites To achieve extremely high thermal neutron fluxes uniformly distributed across a patient's head, for example, a hemispherical geometry is used in one embodiment of the invention. This unique geometry arranges fast neutron sources in a circle around a moderator whose radial thickness is optimized to deliver a maximum thermal neutron flux to a patient's brain. This embodiment produces a uniform thermal neutron dose within a factor of 1/20th of the required fast neutron yield and line-voltage input power of a conventional planar neutron irradiation system. This arrangement permits using a relatively safe deuterium-deuterium (DD) fusion reaction (no radioactive tritium) and commercial high voltage power supplies operating at modest powers (50 to 100 kW). FIG. 2 is a cross sectional view of a hemispheric neutron irradiation system 36 according to one embodiment of the invention. Multiple fast neutron generators 68 surround a hemispheric moderator 34, which in turn surrounds the patient's head 26. Titanium targets 52 are distributed around the perimeter of the hemispheric moderator 34. Surrounding the moderator 34 and the fast neutron generators 68 is a fast-neutron reflector 44. In the moderator 34, moderating material such as 7LiF, high density polyethylene (HDPE), and heavy water are shaped in a hemisphere that is shaped around the head of the patient. The optimum thickness of the hemispheric moderator for irradiation purposes is dependent upon the material's nuclear structure and density. FIG. 3 shows a perspective view of a patient 58 on a table 54 with the patient's head inserted into hemispheric irradiation system 36. The patent 58 lies on the table 54 with his head inserted into hemispheric moderator 34. Surrounding the moderator is neutron reflecting material 44, such as lead or bismuth. Referring again to FIG. 2, fast neutrons 22 are produced by fast neutron generators 68. Generators 68 are composed of titanium targets 52 and ion sources 50. Ion beams are produced by ion sources 50 and accelerated toward titanium targets 52 which are embedded in hemispheric moderator 34. A DD fusion reaction occurs at the target, producing 2.5 MeV fast neutrons 22. The fast neutrons 22 enter the moderator 34 wherein they are elastically scattered by collisions with the moderator atom's nuclei. This slows them down after a few collisions to epithermal neutrons 24 energies. These epithermal neutrons 24 enter the patient's head 26 wherein they are moderated further to thermal neutron 30 energies. These thermal neutrons 30 are then captured by boron-10 nuclei at the cancer site, resulting in a fusion event and the death of proximal cancer cells. Fast neutrons 22 are emitted isotropically from titanium target 52 in all directions. Outwardly traveling fast neutrons 42 are reflected back (reflected neutron 48) by fast neutron reflector 44, while inwardly traveling fast neutrons 40 are moderated to epithermal energies and enter the patient's head 26, where further moderation of the neutrons to thermal energies occurs. A shell of protective shielding 56 is also shown in FIG. 2. In some embodiments, this may be necessary for shielding both the patient and the operator from excessive irradiation due to neutrons, x-rays and gamma radiation. The shielding can be made of a variety of materials depending upon the radiation components one wishes to suppress. In some embodiments, fast neutron reflector 44 is made of lead or bismuth. The fast neutron reflector also acts as a shielding means to reduce emitted gamma rays and neutrons from the hemispherical neutron irradiation system 36. As one skilled in the art will realize, gamma-absorbing or other neutron reflector means can be placed in layers around the hemispherical neutron irradiation system 36 to reduce spurious and dangerous radiation from reaching the patient 58 and the operator. Hemispheric moderator 34, fast neutron reflector 44 and head 26 act together to concentrate the thermal neutrons in the patient's head. The patent's head and the moderator 34 act in concert as a single moderator. With a careful selection of moderating materials and geometry, a uniform dose of thermal neutrons can be achieved across the patient's head and, if a boron drug is administered, a large and uniform therapeutic ratio can be achieved. The invention gives a uniform dose of thermal neutrons to the head while minimizing the fast neutron and gamma contributions. The required quantity of fast neutrons to initiate this performance is reduced compared to that of prior art planar neutron irradiation systems (see FIG. 1). A cross section perspective view of the hemispheric neutron irradiation system 36 in an embodiment of the invention is shown in FIG. 4. This cross-section view is of a radial cut directly through the patent's head 26 and hemispherical neutron irradiation system 36. As shown in this embodiment, ten fast-neutron generators 68 composed of ion sources 50 with titanium targets 52 are radially surrounding the hemispheric moderator 34 and the patient's head 26. The titanium target 52 in this embodiment is a continuous belt of titanium surrounding the moderator 34. The titanium targets can also be segmented, as was shown in FIG. 2. The ion sources in this embodiment are embedded in fast neutron reflector 44. There are a number of materials one could select for the moderator 34 to achieve maximum thermal neutron flux at the patient's head 26. The performance of HDPE, heavy water (D2O), graphite, 7LiF, and AlF3 was analyzed using the Monte Carlo Neutral Particle (MCNP) simulation. In general, there is an optimum thickness for each moderator material that generates the maximum thermal flux at the patient's head (or other body part or organ). The thermal neutrons/(cm2-s) was calculated for these materials as a function of moderator thickness d3, where d4=25 cm, and fast neutron reflector 44 is d1=50 cm thick and is made of lead. As in all our calculations, the combined fast neutron yield striking the area from all the fast neutron generators 68 is assumed in the MCNP to be 1011 n/s. The optimum thickness, range of thicknesses and maximum thermal neutron flux (E<0.5 eV) are given in Table I for various moderator materials. These are approximate values given to help determine the general dimensions of the moderator. TABLE IModerator ThicknessModeratorOptimumRange of thicknessMaximum FluxMaterialThickness d3 (cm)d3 (cm)(n/cm2-sec)HDPE 6 4-10 7 × 108D2O15 9-25 2 × 108Graphite2019-20 9 × 1077LiF2520-30 3 × 107AlF33020-401.5 × 107 The calculation of the therapeutic ratio is also important and depends upon the organ in question (brain, liver) and the body mass of the patient. Although HDPE gives the highest flux, it gives a lower therapeutic ratio compared to 7LiF. The designer is expected to do calculations similar to this to determine the optimum geometry for the neutron irradiation system. The MCNP simulation was used to determine the delivered dose and therapeutic ratio to the patient 58 and compare it to a planar neutron irradiation system. In one simulation, moderator 34 is composed of 7LiF whose thickness is d3=25 cm. The inner diameter of the moderator (hole for head) is d4=25 cm. The spacing between hemispheric fast neutron reflector 44 and hemispheric moderator 34 is d2=10 cm. The head is assumed to be 28 cm by 34 cm. Fast neutron reflector 44 is made of d1=20 cm thick lead in one embodiment. Thicker values of d1 increase the tumor dose rate. At a thickness of 10 cm, the tumor dose rate is about one-half the value at a thickness of 50 cm. Fast neutron generators 68 are assumed to emit a total yield of 1011 n/sec. The combined titanium targets 52 give a total neutron emission area of 1401 cm2. In the MCNP simulation BPA (Boronophenylalanine) was used as a delivery drug. The concentration of boron in the tumor was 68.3 μg/gm and in the healthy tissue was 19 μg/gm. The calculated neutron dose rates in Gy-equivalent/hr are plotted in FIG. 5 as a function of distance from the skin to the center of the head. The calculated dose rates are comparable to those used for gamma radiotherapy, typically 1.8 to 2.0 Gy per session. For the same dosage, at a rate of 3 Gy-equivalent/hr, the session length would be from 30 to 40 min. long. These session times are considered reasonable for a patient to undergo. For this simulation, the therapeutic ratio for the hemispherical neutron irradiation system is plotted in FIG. 6 as a function of distance from the skin to the center of the skull. The therapeutic ratio is defined as the delivered tumor dose divided by the maximum dose to healthy tissue. A therapeutic ratio of greater than 3 is considered adequate for cancer therapy. The conventional planar neutron irradiation system requires larger fast-neutron yields (1012 to 1013 n/s) to achieve equivalent dose rates and therapeutic ratios. In FIG. 5, a planar neutron irradiation system 14 of FIG. 1 is compared with that of a hemispheric neutron irradiation system 36 (FIGS. 2, 3, 4) in one embodiment of the present invention, using the same source of fast neutrons (1011 n/s). As can be seen from FIG. 5, the hemispherical neutron irradiation system (called radial source in FIG. 5) achieves a dose rate of about a factor of 20 over that of the conventional planar neutron irradiation system 14. The planar geometry needs a fast neutron source of 2×1012 n/s to achieve the same results. Indeed, if a DD fusion generator is used, then the planar source requires a factor of 20× increase in wall-plug power or 2.0 MW, a prohibitively large power requirement. In addition, as can be seen from FIG. 5, over a ±5 cm distance across the head center, hemispheric neutron irradiation system 36 has less than a 10% variation in dosage. A uniform dose rate is crucial for the treatment of GBM, where we want to maintain a maximum therapeutic ratio and tumors may have distributed themselves across the brain. Hemispherical neutron radiation system 36 in embodiments of the invention also gives a more uniform therapeutic ratio (FIG. 6) across the brain. The ratio is more uniform for the radial source and requires only 1/20th of the fast neutron yield of the planar source (FIG. 1). Other materials can be used for hemispheric moderator 34 in alternative embodiments. As those skilled in the art will know, high density polyethylene (HDPE), heavy water (D2O), Graphite and 7LiF can also be used. In addition, combinations of materials (e.g. 40% Al and 60% AlF3) can also be used. Different thicknesses d1 of moderator can be used to optimize the neutron flux and give the highest therapeutic ratio. The term “neutron generator or source” is intended to cover a wide range of devices for the generation of neutrons. The least expensive and most compact generator is the “fusion neutron generator” that produces neutrons by fusing isotopes of hydrogen (e.g. tritium and deuterium) by accelerating them together using modest acceleration energies. These fusion neutron generators are compact and relatively inexpensive compared to linear accelerators that can produce directed neutron beams. Other embodiments depend upon the selection of the plasma ion source that is used to generate the neutrons at the cylindrical target. These are (1) the RF-driven plasma ion source using a loop RF antenna, (2) the microwave-driven electron cyclotron resonance (ECR) plasma ion source, (3) the RF-driven spiral antenna plasma ion source, (4) the multi-cusp plasma ion source and (5) the Penning diode plasma ion source. All plasma ion sources can be used to create deuterium or tritium ions for fast neutron generation. Cylindrical Irradiation System for the Liver and Other Cancer Sites. FIGS. 7A and 7B shows another embodiment of the invention which uses a cylindrical geometry to irradiate other organs and parts of patient 58, such as the liver 76. FIG. 7A is a cross sectional view of cylindrical neutron irradiation system 62 and FIG. 7B is a perspective view of the same embodiment. In this embodiment eight fast-neutron generators 68 surround a cylindrical moderator 46. These generators 68 all emit their fast neutrons at the surface of the moderator. A cylindrical fast neutron reflector 44 surrounds the cylindrical moderator 46. As in the case of the hemispheric moderator 34, the cylindrical moderator 62 can be composed of well-known moderating materials such as 7LiF, high density polyethylene (HDPE), and heavy water. These are shaped in a cylinder that surrounds the patient. The optimum thickness of the cylinder moderator for neutron capture purposes is dependent upon the material nuclear structure and density. In this embodiment, fusion neutron generators are used to supply the fast neutrons. Fast neutron generator 68 is composed of a titanium target 52 and an ion source 50 as before. The titanium targets are contiguous to the cylindrical moderator 46. Ion beams 60 are accelerated using a DC high voltage (e.g. 100 kV) to the titanium target 52 where fast neutrons are produced from the DD fusion reaction. The fast neutrons are emitted isotropically from the titanium targets 52 on the moderator, some moving out to the fast neutron reflector 44 and others inwardly to be moderated immediately to epithermal or thermal energies. Those reflected come back into the cylindrical moderator 46 where they are moderated to epithermal and thermal energies, making their way finally to the patient 58. Cylindrical neutron irradiation system 62 permits uniform illumination of a section of the patient's body (e.g. liver) as compared to the conventional planar neutron irradiation system. In the case of the brain, the body itself acts as part of the moderation process, thermalizing epithermal neutrons coming in from cylindrical moderator 46. As one skilled in the art will realize, other cancers, such as throat and neck tumors, can be effectively irradiated by a hemispherical neutron irradiation system such as system 36. The thickness and material content of the moderator can be adjusted to maximize the desired energy of the neutrons that enter the patient. For example, for throat and neck tumors, the moderator can be made of deuterated polyethylene or heavy water (D2O) to maximize thermal neutron irradiation of the tumor near the surface of the body. For deeper penetration of the neutrons one might make the moderator out of AlF3, producing epithermal neutrons. These would be optimum for reaching the liver and producing uniform illumination of that organ. Segmented Moderator In yet another embodiment, fast neutron sources with segmented moderators may be individually moved to achieve a uniform dose across the liver or other cancer site. This geometry produces a uniform thermal neutron dose with a factor of between 1/10th and 1/20th of the required fast neutron yield and line-voltage input power of previous linear designs. This again permits the use of the relatively safe deuterium-deuterium (DD) fusion reaction (no radioactive tritium) and off-the-shelf high voltage power supplies operating at modest power (≤100 kW). A segmented neutron irradiation system 70 in an embodiment of the invention is shown in FIG. 8. Ten fast neutron generators 68, each with a wedge-shaped moderator 74, surround the patient 58. The exact shape of each moderator can vary and can be of other geometries. Each generator and moderator pair can be moved independently of the others to achieve uniformity of the neutron flux across the liver, organ, or body part. In between the wedge-shaped moderators 74 more moderating material (“filler moderating material” 72) is inserted, forming a large single moderator. The “filler” moderating material 72 can be heavy water or powered moderating materials such as AlF3. Pie shaped fillers of moderating material can also be fitted into the spaces between the wedge-shaped moderator 74. Since neutrons scatter easily, there can be some space between the wedge-shaped moderators 74 and the pie shaped fillers without undue loss of neutron moderating efficiency. The neutron yield from and the position of each fast neutron generator 68 can be adjusted to achieve uniformity across the liver or body part. The position and the neutron yield of the generator can be varied to achieve the desired radiation dose at a particular location in the patient's body. Since the cancer can be located in any part of the body, this benefit can be particularly useful for optimizing the dose at the cancer site. Surrounding the entire fast neutron/moderator system is a cylindrical fast neutron reflector 44. Fast neutrons are produced by the fast neutron generators 68 and enter the moderators 74 where they are elastically scattered by collisions with the moderator atoms' nuclei, slowing them down after a few collisions to epithermal energies. As in the other embodiments, these epithermal neutrons enter the patient 58 and liver 76, wherein they are moderated further to thermal neutron energies. The invention in various embodiments provides a uniform dose of thermal neutrons to the liver, organ or body part while minimizing fast neutron and gamma contributions. The required number of fast neutrons (e.g. 2×1011 n/s) to initiate this performance is again reduced compared to that (e.g. 2×1013 n/s) needed for the planar neutron irradiation system of the prior art. Another embodiment of the segmented design is shown in FIG. 9. The shape of the neutron irradiation system 78 is elliptical, with six sources of fast neutrons shown as distributed targets embedded in the inside elliptical moderator 96. Fast neutrons 22 are emitted isotropically in all directions. Those fast neutrons 22 moving outwardly are reflected back (see arrow 48) by fast neutron reflector 44, while fast neutrons traveling inwardly 22 are moderated to epithermal energies and enter the liver 76, where further moderation of the neutrons to thermal energies occurs. The inside elliptical moderator 96, outside elliptical moderator 98, reflector 44 and patient's body 58 act together to moderate and concentrate the thermal neutrons into the patient's liver 76. With a careful positioning of the moderators and fast neutron sources 90, 92, 94, a uniform dose can be achieved across the patient's liver, and, with a boron drug administered to the tumor, an excellent therapeutic ratio can be achieved. Elliptical neutron irradiation system 78 in FIG. 9 is a simplified cross-sectional view of the patient 58 inside the elliptical moderator 96. This cross-section view is of a radial cut directly through the patent's torso and the moderator and fast neutron generator system. To maintain visual simplicity, only the titanium targets are shown and not the ion sources. Thus, six fast-neutron sources are represented by three flat titanium targets 90, 92, 94. The rest of the fast neutron generator is not shown. Other components (e.g. plasma ion source) are neglected in the analysis. The wedge-shaped moderators 74 (used in FIG. 8) are also not shown in FIG. 9. For a simple simulation of the neutron irradiation system, the targets 90, 92, 94 are the sources of the fast neutrons and are arranged in an elliptical material 96 (e.g. AlF3, LiF). The effect of the moderating material 96, the fast neutron reflector 44 and the patient's body 58 were calculated using a Monte Carlo N-particle (MCNP5) transport code to determine how fast the neutrons were converted to thermal neutrons in the neutron irradiation system. Dosage calculations were made along a central axis of the liver. The fast neutron sources (titanium targets) are 2 cm×2 cm in area, each producing 1011/N n/s, where N is the number of sources. The human body 58 dimensions are 35.5 cm along the major axis and 22.9 cm along the minor axis. The inner elliptical moderator 96 is made of 7LiF and 10 cm thick, while the outer moderator 98 is made of AlF3 and 40 cm thick. The fast neutron reflector 44 is made of lead 50 cm thick. Boron-10 concentration is 19.0 μg/g in the healthy tissue and 68.3 μg/g in the tumor. The six sources are located in cms at: (−15,18.06,0) (−15,−18.06,0) (−17,17,0) (−17,−17,0) (0,15.85,0) (0,−15.85,0). These measurements are made along the axis of the liver 76 from the point (−15,0,0) to (−5,0,0). In the x-direction, the first two sources 90 are centered about the left edge of the liver shown in FIG. 9, the two sources 92 are centered about the edge of the body, and the third two 94 are located above and below the origin. The origin is shown in FIG. 9 as a small cross + at the center of the body in the plane of the liver. FIG. 10 shows the therapeutic ratio for a large single dose, and the therapeutic ratio for multiple small doses (where the photon dose to healthy tissue is not included) plotted as a function of distance along the axis of the liver. The photon dose can be neglected if there is some amount of time between doses. Many of the body's healthy cells can self-repair and recover between doses. The expected therapeutic ratio is between these two curves when there is fractionation into multiple doses. In this simulation, BPA was again used as the delivery drug with the concentration of boron in the tumor at 68.3 μg/gm and in the healthy tissue at 19 μg/gm. FIG. 11 indicates that the goal of having an extremely uniform dosage to the tumor has been achieved, with about ±6% variation along the x-dimension. The calculated dose rates are comparable to those used for gamma radiotherapy, typically 1.8 to 2.0 Gy-equivalent per hour if we increase the total neutron yield to 2×1011 to 3×1011 n/s. Thus, at approximately 2×1011 to 3×1011n/s it is possible to obtain a therapeutic ratio and uniform dosage to a tumor. Approximately 10 to 20 treatments of 30 to 40 minutes would be required, with a good therapeutic ratio, uniformity of dosage, and the opportunity for healthy tissue repair between treatments. Once again, the planar neutron irradiation systems require high fast neutron yields to drive them. In one prior art system known to the inventors a fast neutron source of 3×1013 n/s is needed to obtain realistic treatment time of ˜1-2 hours. Using a D-T neutron source with a yield 1014 n/s, acceptable treatment times were obtained (30 to 72 minutes with single beam and 63 to 128 minutes with 3 beams of different direction). But these are impossible yields to achieve with realistic wall plug powers. Instead of 50 to 100 kW for the hemispheric and cylindrical neutron irradiation systems, it would take a minimum of 0.5 MW to achieve adequate yield for the planar geometry with a DT generator. These are high powers for clinics and hospitals. As one skilled in the art knows, other cancers, such as throat and neck tumors, can be effectively irradiated by the neutron irradiation system. The thickness and material content of the moderator can be adjusted to maximize the desired energy of the neutrons that enter the patient. For example, for throat and neck tumors, the moderator can be made of deuterated polyethylene or heavy water (D2O) to maximize thermal neutron irradiation of the tumor near the surface of the body. For deeper penetration of the neutrons one might make the moderator out of AlF3, producing epithermal neutrons. These would be optimum for reaching the liver and producing uniform illumination of that organ. Modular Generators Introduction As is shown in FIGS. 8 and 9, multiple modular generators may be encased in moderator material and may be arrayed to maximize thermal neutron flux at a cancer tumor location. Fast 2.5 MeV neutrons must be slowed (moderated) to energies (usually epithermal) that will penetrate to the cancer site without too many neutrons being lost in their travel to the cancer via capture by healthy tissue. These modular generators act as independent neutron sources and each may be optimized by adjustment of each individual beam's energy, direction and intensity. The modular generators can be arranged to fit a site in a particular subject's component location and structure. This is true also for cancer tumor location. The energy of the neutrons can also be adjusted by adding or subtracting moderator material. This can be done more easily than with a single beam LINAC or reactor, which usually has a fixed beamline that is integral to the neutron source. In the prior art some adjustment can be made, but the DD fusion generator in embodiments of the invention, being much smaller, can have more degrees of freedom in direction, intensity and moderation. This has an added benefit of aiding physicians in tailoring neutron radiation to the patient's cancer. Comparison to Linear Accelerators and Reactors. Modular generators in various embodiments of the present invention may also form and be part of the mechanical structure of a cancer irradiation system. This has an added benefit of moving the neutron sources as close as possible to the cancer site and the diseased body part, resulting in efficient use of the neutron source. The neutrons are being emitted in a 4π solid angle from the modular generators, so the closer to cancer site, the more of the fast neutron flux is being utilized. Linear accelerators (LINACs), which are somewhat collimated, are further from the cancer site and cannot provide this advantage. Compared to a linear accelerator, which can be several meters long or longer and may include large microwave power sources, the DD fusion sources in embodiments of the invention are less than one meter long and comprise compact microwave sources that can either be solid state microwave sources or small, inexpensive, single microwave oven magnetrons. The accelerator structure in embodiments of the invention is compact and includes a pre-moderator 118 that adds only from 5-10 cm of High-Density Polyethylene (HDPE) or 15-20 cm of polytetrafluoroethene (PTFE) Teflon to produce a first stage of neutron beam tailoring. The pre-moderator in these embodiments is an integral part of each modular generator, as is taught below with reference to several figures. In alternative embodiments other pre-moderator materials can be used such as AlF3, MgF2, 7LiF, and Fluental (trade name). Smaller, Nontoxic, Less Complex Targets for Neutron Production The modular DD fusion generator 118 in embodiments of the present invention uses a small titanium target (e.g. a 5 cm diameter disk of titanium backed by water-cooled copper fins) to produce neutrons. The target is supported directly on the pre-moderator, which is an integral part of the apparatus in this application, termed a modular generator. Linacs and other methods in the conventional arts use larger or toxic targets that require complex cooling and rotation. For example, the neutron source used by Neutron Therapeutics has a 2.6 MeV electrostatic proton accelerator and a rotating, solid lithium target for generating neutrons. In that prior-art process the Lithium becomes radioactive and toxic, and when exposed to air, it disintegrates. This prior art source has a large target chamber housing a large Li disk which is rotated in a powerful 2.8 MeV proton beam produced by a large accelerator. The Lithium wheel is roughly 2 meters in diameter and has been divided into pie-shaped sections that are removed by mechanical robotic means. In embodiments of the present invention, the Ti target is a relatively small diameter (˜5 cm) and is typically attached with 6-8 screws to the pre-moderator block and is sealed to the block with a Viton “O” ring. The Ti targets in embodiments of the invention can be easily manually removed and replaced. They also have a long lifetime and have been tested for over 4000 hours with no failures. Nuclear reactors are large structures with a substantial amount of shielding (water and concrete) and cooling systems to maintain the hot reactor core. Reactors provide primarily thermal neutrons that must be raised up in energy using an energy multiplier, and then the neutron beam must be improved to IAEA standards to produce epithermal neutrons with minimal gamma radiation. Optimizing Neutron Energy for Penetration and Minimum Damage to Healthy Tissue For tumors at depths in a subject of 3 cm or more, a goal for the moderator is to provide a neutron beam that has its energy clustered about 10 keV at the skin, in order to provide sufficient energy to penetrate a minimum of several centimeters into a human target while avoiding higher energies that are more damaging to human tissue. High conversion to epithermal energies occurs in HDPE at a thickness of approximately 5 cm, but it also produces a high yield of thermal neutrons and 2.2 MeV gammas that can damage the healthy tissue at the skin. Modular Generators In embodiments of the present invention modular generators are very important components. The modular generator combines multiple functions that were separate functions in the prior art. These integrated functions include both neutron production and beam tailoring. FIG. 12A is a perspective view of an individual modular generator 118 in an embodiment of the invention. FIG. 12B is a cross section of the modular generator 118 of FIG. 12A taken along an axis of an acceleration chamber 100 for ion beam generation and containment, and at a right angle to the axis of a turbo vacuum pump 124 that is part of the modular generator 118. FIG. 12C is a cross section of the modular generator 118 of FIG. 12A taken along the axis of the acceleration chamber 100, and along the axis of the turbo vacuum pump 124, at a right angle to the section of FIG. 12B. Each modular generator 118 can operate independently of the other modular generators and each possesses all required components to generate neutrons. Further, the various modular generators may have pre-moderators shaped to engage other building blocks of a project, such as adjacent generators or spacing moderators, as is described in enabling detail below. Viewed as in FIGS. 12A, B and C, each modular generator 118 comprises a pre-moderator 108 that is made of material known to moderate energy of energetic neutrons. In most embodiments the pre-moderator is a solid block of material, with a rather complicated shape for certain purposes. Modular generator 118 has three key elements: (1) a deuterium ion source 102, (2) an acceleration chamber 100, through which deuterium ions may be accelerated, and (3) a titanium target 106 (shown in FIGS. 12B and 12C) that is bombarded by the deuterium ions to produce high-energy neutrons. The deuterium ion source 102 has an attached microwave source 160, and microwave slug tuners 172, connected by a cable 178. Deuterium gas is leaked slowly into a plasma ion chamber 174 at the upper end of the acceleration chamber, where microwave energy ionizes the gas, creating deuterium D+ ions. The gas is ionized by microwave energy, and Deuterium (D+) ions are created and accelerated out through an ion extraction iris 138 into acceleration chamber 100, and through an electron suppression shroud 180 which deflects back-streaming electrons from being accelerated back into the plasma source, which could damage the apparatus. Electrons are being created by collisions of the D+ ions in the deuterium gas that are being created in the acceleration chamber. The deuterium ions are positively charged, and target 106 is negatively charged to a level of from 120 kV to 220 kV, and the D+ ions are strongly attracted to negatively biased target 106. Acceleration chamber 100 is connected to a turbo vacuum pump 124 that provides a modest vacuum in one embodiment of about 10−6 Torr, minimizing scattering of the D+ ions as they travel from the extraction iris 138 to the target 106. Titanium target 106 is positioned in a primary electrically insulating well 181 at the bottom of the chamber embedded into the pre-moderator material, which may be UHMW, HDPE or Teflon, of the pre-moderator 108. There is further a secondary electrical insulating well 182 surrounding the primary electrical insulating well. The surface of the moderator material in the primary and secondary electrical insulating wells may be seen as a corrugated insulator causing any surface charge to follow a curved path taken in any direction. The purpose is to provide a very long surface path to prevent electrons from traveling from the target to acceleration chamber 100 wall or any grounded element, and to avoid surface electrical breakdown or flashover in that surface path. As those skilled in the art know, the wells form an electrical insulating path. Additional corrugations or wells can be added to lengthen the path. Pre-moderator 108 has a high voltage bus bar 122 and fluid cooling channels 120 to and from the target. The high voltage is introduced via a high voltage receptacle 130 which is connected to the high voltage bus bar. Pre-moderator 108 acts as a HV insulator and as a mechanical support for the target 106 at a high negative bias. The pre-moderator 108 has metal cladding 140 at ground potential to minimize high voltage breakdown through the pre-moderator plastics. When in operation the D+ ions in the ion beam are attracted to the titanium target 106, where fast (2.5 MeV) neutrons are produced in a resulting DD fusion reaction. FIG. 13A illustrates an assembly of six modular generators 118, wherein pre-moderators 108 are spaced apart by spacers 128 which are also made of moderator material. FIG. 13B shows the arrangement of FIG. 13A in perspective. FIG. 13C shows the arrangement of FIG. 13B with one modular generator 118 removed from the assembly. FIG. 13D is a more diagrammatic illustration showing an arrangement in which modular generators may be mounted on translation and rotation mechanisms to be positioned to maximum irradiation of a cancer site. As is shown in FIGS. 13A-D the modular generators in embodiments of the invention may be arranged in an array to form a complete and moveable system of irradiating neutron sources with pre-moderators. For example, as shown in FIG. 13A-C, in the simplest configuration of the array, the modular generators may form a circle around a human torso or body part. The modular generators can be moved into three dimensional arrays around the subject to maximize neutron flux to a cancer site 148 that may not be centered on a body part 146, illustrated as a human brain in FIG. 13D. Thus, depending upon body contour, shape and size, and cancer location and distribution, the modular generators may be moved to adapt to the shape and tumor location in order to maximize the dose to the cancer and to minimize the dose to the other body parts. Referring to FIG. 13D, rotation 150 and translation 151 of the modular generator 118 can be achieved with electrical motors attached to the modular generator 118. Seven Functions of the Pre-Moderator Because the titanium target is on the pre-moderator (first stage of moderation), fast neutrons coming from the target immediately enter the pre-moderator and quickly moderated to thermal or epithermal energies. The pre-moderator also provides mechanical support, high voltage supply and cooling fluid transport to the titanium target. Exemplary pre-moderator materials that may accomplish this are Teflon and HDPE. Both Teflon and HDPE are excellent high voltage dielectrics which can also support a HV bus bar 122 and water channels 120 to be used to transport HV and the cooling fluids to the Ti target, as shown in FIG. 12C. As shown in FIGS. 12A, B, C a single generator 118 consists of an acceleration chamber 100, an ion source 102 emitting deuterium ions, a titanium target 106 and a pre-moderator 108. Pre-moderator 108 also provides a function of being a high voltage insulator for high voltage bus bar 122 that delivers high voltage (e.g. 80 kV to 300 kV)) to titanium target 106, and a water channels 120 that deliver cooling fluid to the titanium target 106. The high voltage is delivered from a high voltage power supply through a standard HV receptacle 130 to the bus bar 122 and then on to the titanium target 106, all of which are mounted in the pre-moderator 108. In various embodiments of the invention the pre-moderator 108 performs seven functions: (1) moderation, (2) mechanical support of the titanium target, (3) cooling fluid transport to the target, (4) high voltage transport to the target, (5) minimum surface flashover, (6) and a portion of a high vacuum container (a wall) with no out gassing (7). These seven attributes permit a substantial reduction of distance and amount of material between the fast neutron source and the patient, thus helping to maintain a maximum neutron flux delivered to the patient. Modular Generators Around a Subject FIGS. 13A-D show how the generators may be arranged. In FIG. 13A, six modular generators 118 form a ring around a secondary moderator 112 and are part of a structure formed by secondary moderator 112, spacers 128, and pre-moderators 108. Pre-moderators 108 and secondary moderator 112 provide the moderation function by slowing the neutrons down to epithermal energies (function #1). These elements also form a mechanical support (function #2) for the entire generator and moderator system. Secondary moderator 112 may also be a separate section attached directly to the modular generator just after the pre-moderator, each separate from the other instead of being in a ring 112 as in FIG. 13A. As shown in FIG. 12B-C, fluid transport (function #3) is supplied through channels 120, which delivers cooling fluid to target 106 to maintain the target at an acceptable operating temperature. Each generator is supplied with a separate cooling fluid input and output, wherein cooling fluid is provided through a connector 132 shown in FIGS. 12A-12C. Thus, the pre-moderator supplies fluid transport (function #4). High voltage is delivered via high voltage bus 122, which passes through pre-moderator 108 (function 4, high voltage transport). HDPE, UHMW and Teflon are excellent insulators and withstand high voltage flashover (function #6). All three may be used in vacuum systems without excessive out gassing and may help maintain the system vacuum (function #7). The achievement of these seven functions provides a very compact and flexible neutron source. The Secondary Moderator Secondary moderator 112 (FIGS. 13A-C) may comprise any one of or a combination of multiple moderator materials that optimize both the maximum flux and neutron energy for maximum dose to the cancer site. Selection (material, size and shape) may be varied depending on depth of the cancer in the subject and a desired dose at the cancer site. The secondary moderator may be D2O (heavy water) for delivery of thermal neutrons to, for example, throat and neck cancers, or a combination of AlF3 and Teflon for delivery of epithermal neutrons to brain tumors. The recommended levels of fast, thermal and gamma emission by IAEA are given in Table I. TABLE 1values in window IAEA Recommended the beam exitIAEA RecommendedBNCT beam port parametersvalueϕepithermal (n cm−2 s−1)~109ϕepithermal/ϕfast>20ϕepithermal/ϕthermal>100Dfast/ϕepithermal (Gy cm2)<2 × 10−13Dγ/ϕepithermal (Gy cm2)<2 × 10−13Fast energy group (ϕfast)E > 10 keVEpithermal energy group1 eV ≤ E ≤ 10Thermal energy group (ϕthermal)E < 1 eV These IAEA recommended values depend upon older drugs, such as p-Boronophenylalanine (BPA) that have been approved for use in humans by the Food & Drug Administration (FDA) for other medical applications. Delivery of higher boron concentrations to a cancer site may depend to some extent on newer drugs to be developed, and may permit lower power, less efficient neutron beams to be used. Since treatment time might also be faster, the neutron beam quality need not be as high. DD fusion generators in embodiments of this invention have relatively low beam flux, thus permitting them to be used for cancer therapy. In some embodiments multiple modular generators may be distributed around a secondary moderator surrounding a central chamber holding a subject for treatment, providing an alternative to a completely integrated multi-ion beam system, and may have particular benefits in some circumstances. Benefits might include (1) an ability to quickly replace a single generator that has failed and needs repair; and (2) an ability to change alignment of the generators relative to one another, the moderator, and the subject. In regard to a subject, alignment of the generators may optimize dose distribution and density of neutrons at a cancer site, while at the same time minimizing spurious radiation, such as gamma rays that might be emitted external to the apparatus, or into healthy tissue of the subject. In the prior art, where reactor and accelerator neutron sources are used, careful attention has been given to achievement of high quality neutron beams to meet the IAEA standards for BNCT developed in 2001 for International Atomic Energy Agency (IAEA) (Current Status of Neutron Capture Therapy (2001) IAEA-TECDOC-1223. In embodiments of the present invention, where multiple modular DD fusion generators are used, these standards may be relaxed. The IAEA specification assumes that there is a single neutron beam that is used for all cancers and body locations. This results in standard values for the three neutron energies (thermals, epithermal and fast neutrons). Moderator and neutron spectral shifters are then designed to achieve these values for a particular fast neutron source as an input specification. This results in designs in the prior art that may not use the available fast neutrons economically and then may waste some of them to achieve the IAEA universal specs. For generators such as the DD fusion source in an embodiment of the present invention, early calculations have indicated that a single DD fusion generator would have difficulty achieving required fast neutron input to the moderating process. So, in embodiments of the invention, the use of multiple generators increases the total fast neutron yield available and allows the moderated dose to be distributed over a larger area of the body, instead of having the beam enter at one location of the body. For example, as shown in FIG. 13D, neutrons n are entering the head from many directions. This permits reduction of thermal neutron flux at any one point on the skin of the head while still achieving adequate epithermal flux to the cancer site. In early prior art reactor BNCT experiments, the thermal neutron flux burned the skin of subjects. When considering neutrons used for a particular cancer it is desirable to direct the maximum flux to the cancer site, and therefore, one must consider the specific cancer that is to be treated. This includes location and depth in the human body. Because of their relatively small size and large neutron yield, the modular generators in the embodiments of the present invention are particularly able to accomplish this by being positioned to maximize their flux at the cancer site. Since in embodiments of the invention generators are placed as close to the patient's body as practical to maximize flux at the cancer site, there is a more holistic problem. There are multiple parameters for each modular generator: (e.g. neutron flux, neutron energy, position relative to the body). What comes out of a single neutron beam pipe (1998 IAEA Standards, Table I) is not the only concern. A body part can now, in new implementations of the invention, be irradiated in all directions, and neutron intensity can be adjusted at each modular generator to achieve better flux and even more optimum neutron energy than with a single beam LINAC or a reactor. The direction of each neutron beam can be adjusted by rotating and displacing each modular generator 118. Each modular generator's yield can be adjusted electronically by varying the accelerator voltage and the ion beam current. Since the moderator size is relatively small and compact compared to the prior art, the neutron spectrum of each modular generator 118 can be adjusted by the selection of different moderator materials and thicknesses. Lowering of Required Beam Quality In embodiments of the present invention the subject's body is bombarded with neutrons from multiple directions. The neutrons can come from all sides of the body part, which minimizes the amount of distance each beam has to transverse. Unwanted neutrons striking the skin are now distributed over a larger area, reducing the skin dose of harmful components (e.g. gammas, and thermal and fast neutrons) per unit area. These components are simply delivered over a larger area of the skin. This permits adjustment of dose at the cancer site to be higher than that achieved with a single beam but with reduction of harmful components over a larger area of the skin. For a single beam case in the prior art, an argument might be made that one can rotate the patient for each exposure, but, due to possible patient movement, the neutrons would not be as accurately placed as in multi-beam embodiments of the present invention. For each placement the patient would have to be carefully re-oriented relative to the single neutron beam, which requires careful placement of the patient. In embodiments of the invention, multiple beam directions and an ability to adjust the neutron flux of each modular generator allow for optimum delivery to the cancer site while reducing harmful components. For example, if the cancer is located in the left lobe of the brain, the neutron flux to the tumor can be adjusted to deliver epithermal neutrons in the direction of that tumor. Since each modular generator neutron flux can be adjusted quickly by varying the accelerator's high voltage or the ion beam current, and by translation and rotation, this can be done easily with delivery determined by a computer program. In the present invention, a control computer monitors the ion beam current, the acceleration voltage and the output neutron yield, which can be automatically adjusted. Small modular generators in embodiments of the invention can make use of new boron drug delivery methods for higher concentrations of boron to the cancer sites. Higher concentrations of boron lower the required neutron dose and require shorter delivery time. Higher boron concentrations to the cancer site permit use of neutron generators with lower neutron yield such as the modular DD fusion generators in embodiments of the present invention. Each modular generator 118 is an independent device capable of producing neutrons independently of the other generators. This allows the total available power, P, to be distributed over N generators, resulting in the heat load being distributed safely without, for example damaging the titanium targets (unlike single target devices using lithium). In one example there are six modular generators, distributing total heat load per titanium target, since the number of neutrons per unit area is fixed by the ion beam power per unit target area. To properly treat a tumor in a subject, a large number of neutrons is required. For reasons of temperature management and stability, DD fusion generators are at present limited to fast neutron yields of less than 4×1010 n/sec. To increase the neutron yield, the number of neutron generators can be increased in embodiments of the present invention. Pre-moderators 108 can be shaped so that larger numbers of modular generators may be fitted around a subject to be treated. In the example shown by FIG. 13A there are six generators arranged equally spaced around a common secondary moderator 112, the subject cavity 116 and the subject 134. Spacing blocks 128, composed of moderator material that may be the same as that of pre-moderator 108 (e.g. Teflon or polyethylene), are placed between each pre-moderator to provide adequate spacing for fitting the subject cavity 118. The wedge angle, α, as indicated in FIG. 12A, on the pre-moderator in FIG. 13A determines the number of modules 118 with pre-moderators 108 that can fit in the circle around the patient and how close the sources may be to the patient. For example, a wedge angle of α=30° for 6 generators and α=22.5° for 8 generators. Moveable Sources with Fluid Moderator One embodiment of a system of modular generators is shown in FIGS. 13A and 13B. In FIG. 13A a plane view of six modular neutron generators 118 fitting into the cylinder (or ring) is shown. In FIG. 13B, a perspective view is shown. The modular generators can also be arranged in other patterns to maximize the dose in particular locations in the subject's body and deliver cancer therapy to selected body organs. In some embodiments of the invention the modular generators may be moved by electric motors and mechanical means to optimized locations to provide the maximum dose to the cancer site and tumor profile as determined by boron bio-distribution test biopsy and pathological analysis, Positron Emission Computed Tomography (PET-CT), Computed Tomography (CT) or magnetic resonance imaging (MRI). One may make use of moderating materials between movable modular generators. For clinical systems there should be moderator material between the modular generators. Ideally the material can quickly position itself to the new location of the modular generators and also be a moderating material. As shown in FIG. 13D, liquid moderator 156 can be used to surround the modular generators 118, acting as a secondary moderator. The moderating material is shown between the movable modular generators. The liquid is contained in an appropriate liquid container. Liquids that also have good moderating properties can be used and are easily displaced by the modular generators when moving. For example, different grades of 3M™ Fluorinert™ Electronic Liquid (e.g. FC-40), which is non-conductive, thermally and chemically stable fluid, can be inserted between generators. Like Teflon it contains primarily fluorine atoms, making it an excellent moderator, and no hydrogen. Stages of Moderation Use of multiple modular generators in embodiments of the invention permits efficient use of modulator material, reducing size of moderator and shielding material and, thus, the reduction and size of the entire system. It also reduces the required flux density of fast neutrons by bringing the neutron sources closer to the patient and directing the limited number of neutrons to the cancer site in a more efficient fashion. The subject's body also becomes part of the equation of the moderating process. The fact that the neutrons are coming from multiple directions reduces local skin dose and localized body dose of healthy tissue. Rather than coming into the body at one location, the neutrons are coming from roughly 360 degrees around the body. Moderation of fast neutrons in embodiments of the invention is a three-step process. In a first step (1) the pre-moderator 108 acts to reduce energy of the fast neutrons in as short a distance as practical with a minimal amount of gamma radiation produced in the process. The pre-moderator also serves as a medium to (2) transport high voltage and (3) cooling fluid to a fast neutron production titanium target 106. Combining these three functions ((1) moderation, (2) fluid transport and (3) high voltage transport) reduces distance and the amount of material between the fast neutron source and the patient, helping to maintain a maximum neutron flux finally delivered to the patient. Partially slowed neutrons can then pass into the secondary moderator 112 which continues the slowing process without undue production of gamma rays from, for example, hydrogen. In the case of small animal models, the selected moderator may be heavy water (D2O). Neutron energy reduction is continued by the D2O without the generation of ˜2.2 MeV gammas that would occur if materials composed of hydrogen were used. For the case of irradiating tumors of depth greater than 3 cm in a human body, the neutrons need to be moderated to epithermal neutron energies. The human body also acts as a partial, final moderator. Thus, the epithermal energy neutrons are slowed further as they move through the body, and finally are slowed to thermal energies at the tumor site. Those skilled in the art will understand that the moderation is a statistical and random process that reduces the neutron energy with a variation or spread of the neutron energies. The process can also result in undesired gamma ray components (e.g. 2.2 MeV gammas from hydrogen capture of neutrons) which damage health cells. In embodiments of the invention, selection of the moderator material depends at least in part upon the desired energy of the neutrons at the body's skin to achieve maximum penetration to the cancer site while reducing (1) excess thermal energy components at the skin, (2) the cost and availability of the moderator material, and (3) harmful gamma ray components. Each generator's energy, yield, direction and moderation can be determined from moderation materials, the generator's voltage and acceleration current. Unlike in the prior art, dimensions of the moderator and content may be quickly changed. In some embodiments of the invention a liquid moderator (e.g. Fluorinert FC40) or a granular (e.g. AlF3) moderator may be used. The modular generators are positioned in the liquid or granular moderator material, where they are free to move by mechanical means quickly between different cancer sites. In the prior art, the moderators and shields are large, massive and usually fixed relative to a single beam reactor or linear accelerator. The patient is usually moved relative to the fixed neutron source. Using liquid or granular moderator materials permits a more efficient reduction of fast neutrons to epithermal energies while minimizing thermals and fast neutrons. Selection of the pre-moderator material is important for optimum neutron beam quality. Generally speaking, beam quality involves minimization of harmful components of radiation that accompany the production of thermal neutrons at the cancer site but also the minimization of the fast and thermal neutrons at the skin surface. In this process gamma rays are produced and, depending upon the cancer site, fast neutrons must be converted to the right energy so that they penetrate the body and deliver thermal neutrons to the tumor site with minimal irradiation of healthy tissue. Moderating the neutrons to thermal energy can result in the skin being damaged. Indeed, the thermal neutron dose to the skin can be larger than the dose to the tumor. The body itself moderates and absorbs the neutrons as they penetrate the body. Selection of the moderator material requires materials that do not moderate the fast neutrons too quickly to thermal energies. Thermal neutrons can damage the skin, and if hydrogen atoms are present in the moderation process, then damaging gamma rays are also produced. Like the moderator, the human body also moderates and absorbs the neutrons. The desired required depth of penetration depends upon the location of the tumor in the body. Simulations show that penetration of thermal neutrons starting at the skin results in penetration depths of 3 to 5 cm before a large fraction of the neutrons are absorbed. Teflon Moderator for Clinical Machine When used as a Pre-moderator, Teflon (PTFE) can satisfy 6 of the 7 functions listed above. Indeed, on several of the attributes Teflon excels. For example, since Teflon does not have atomic hydrogen, gamma production is avoided, whereas the use of HDPE does have hydrogen and, therefore maximizes the thermal neutron moderation with and added 2.2 MeV gamma ray component. Selection of HDPE as the pre-moderator material results in production of thermal neutrons in a short distance from the Ti target, whereas the use of Teflon results in a slower rate of neutron energy reduction from 2.5 MeV permitting the production of epithermal neutrons for deeper penetration into the human body and no 2.2 MeV gammas. Teflon can have a minimum high voltage in which surface arcs (flashovers or surface discharges) momentarily short out the high voltage, and lead to damage to the Teflon surface and possibly damage to the high voltage power supply. This is primarily a materials problem and not a structural problem (shape of the accelerator and Teflon shape and structure). Surface discharge along solid insulators in a vacuum in high voltage devices determines the maximum voltage between an anode and a cathode. The voltage hold-off capability of a solid insulator in vacuum is usually less than that of a vacuum gap of similar dimensions. O. Yamamoto et. al (Yamamoto, O; Takuma, T; Fukuda, M; Nagata, S; Sonoda, T “Improving withstand voltage by roughening the surface of an insulating spacer used in vacuum,” IEEE TRANSACTIONS ON DIELECTRICS AND ELECTRICAL INSULATION (2003), 10(4): 550-556) has studied a simple and reliable method to improve surface insulation strength of a dielectric such as Teflon, PMMA, and SiO2 by roughening the surface of the dielectric. Some experimental results have revealed that in a vacuum, charging along the surface of an insulating spacer precedes the flashover. The charging takes place through a process in which electrons are released from a triple junction where the cathode, insulator and vacuum meet, and propagate toward the anode, causing a secondary emission electron avalanche (SEEA) along the insulator surface. The dielectric (e.g. Teflon or HDPE) can hold charge like a battery or capacitor and then release it along the surface. This limits the use of plastics such as Teflon and HDPE as insulators and moderators inside the vacuum chamber of the neutron generator's acceleration chamber 100. For short distances across Teflon (10 mm), Yamamoto found that roughing the surface (e.g. with sandpaper or sandblasting) affects the charging of various plastics (such as Teflon and HDPE), which decreases as roughness increases. Yamamoto used roughness up to 37.8 μm but had used lower voltage gradients and smaller dielectric thicknesses (10 mm). Studies in embodiments of the present invention find that larger surfaces (distances e.g. 8 inches) of Teflon can be roughened with roughness values of 5 microns and greater and achieve high voltages of 150-220 kV for distances greater than ˜2 cm without flashover. More importantly, the roughing method gives higher insulation strengths without time-consuming conditioning previously used. This provides a significant advantage and makes generators in embodiments of the present invention operational more quickly. Depending on maximum field strength required, conditioning by the roughing process could take minutes or days. Below 1 MV m−1, the conditioning process is relatively fast. Between 1 and 10 MV m−1, the conditioning process takes longer. The best way to monitor how conditioning is going is to record the number of transient discharges (or sparks) per hour. At very high fields the arc rate might never get better than a few arcs per hour. A tolerable arc rate depends on the application. If no high voltage breakdown (arcs) can be tolerated, then the system must first be conditioned to a higher field, and then when the voltage is reduced to the operating level the arc rate drops almost to zero. For very high field strengths above 10 MV m−1, it is very difficult to condition the electrodes to give an arc rate of zero. The electrode shape and material composition becomes very important at these field levels. The Importance of the Human Body in the Moderation Process The human body acts as a moderator to reduce the epithermal neutrons to thermal energies at the cancer site. The amount of neutron energy reduction by the human body depends at least in part upon the depth of the tumor in the body. This determines the maximum neutron flux for delivery to the patient. The desired reduction of the neutron's energy will depend upon the depth of the tumor in the human body. For example, with throat and neck cancers the reduction of the neutron energy to thermal energies is desired for maximum dose to the cancer site. For small animal models, thermal energies are also desired. Dimension in the body from the skin (epidermis) to the cancer site can vary, requiring the neutron energy to be large enough for penetration to the cancer while still primarily at thermal energies, permitting capture by the boron and the destruction of cancer cells. For small animal models or skin cancer in humans, the neutrons can be at thermal energies. For cancers at deeper depths in the body, epithermal neutrons (0.025 to 0.4 eV) can be used. For deep tumors in the torso, such as, for example, pancreatic tumors, epithermal neutrons are required. Pancreatic tumors are deep in the torso and require epithermal neutrons at entrance to the body to penetrate to the tumor. Moderation of the epithermal neutrons occurs as they pass though the body. Simulations in various embodiments show that there are materials at the right thicknesses, such as Teflon, 7LiF and AlF3, which produce the epithermal neutrons that penetrate the body and thermalize by the time they reach the depth of the tumor with a maximum neutron flux. In embodiments of the invention, this occurs while minimizing production of thermal neutrons at the skin. Shape of a Clinical Machine to Match a Human Body The shape of the patient's chamber in a machine may be contoured to fit the human body part to maximize radiation to the cancer site. The shape depends upon the body part to be irradiated and the location of the tumor. As shown in FIG. 13D, for glioblastoma 148 (brain cancer), modular generators 118 may be arranged in a close ring around the head 146 that maximizes neutron flux to the cancer site 148 in the brain. The intensity of each generator can be varied to achieve maximum thermal neutrons to the tumor while minimizing the dose to healthy tissue. As discussed above, applications in embodiments of this invention permit control of the distance of each generator from the cancer site. The cancer site may be mapped using radiographic means (CT scans) and/or MRIs. A treatment planning protocol can then be determined for the optimum use of the clinical neutron source. The intensity of the neutrons coming from each neutron generator can then be varied and the location of each individual generator can be optimized. As shown in FIG. 13 D, an improvement of the moderator surrounding the modular generators is to suspend or surround the modular generators with a liquid 156 that does not contain hydrogen (a gamma producing source), but has modest atomic-number atoms like Fluorine, Carbon or Nitrogen. Various kinds of Fluorinert (tradename), FC-70 or FC-40, or FC3839 can be used. The fluid may be put between the modular generators and by mechanical means each modular generator can move independently of the other generators to a certain extent. This fluid can also absorb some heat from modular generators. As shown in FIG. 13 D, an improvement of the moderator surrounding the modular generators is to suspend or surround the modular generators with a liquid 156 that does not contain hydrogen (a gamma producing source), but has modest atomic-number atoms like Fluorine, Carbon or Nitrogen. Various kinds of Fluorinert (tradename), FC-70 or FC-40, or FC3839 can be used. The fluid may be put between the modular generators and by mechanical means each modular generator can move independently of the other generators to a certain extent. This fluid can also absorb some heat from modular generators. Generator Alignment In embodiments of the present invention each stand-alone generator, as seen in FIG. 13D, for example, may be positioned and aligned to give a maximum flux and neutron distribution at the cancer site. Each generator is small enough in size and weight that the generators may be mechanically moved and positioned so that optimum neutron flux at the cancer site is achieved, depending upon the cancer's location and distribution. The generators may be arranged around a moderator whose radial thickness is optimized to deliver a maximum thermal neutron flux to the cancer site. Depending upon the body part being irradiated, the geometry can be circular or elliptical. By selecting the moderating material and radial thickness one can deliver thermal neutrons to the cancer site. FIG. 14A shows an on-axis view of an exemplary clinical neutron source using multiple modular generators 118 for BNCT of a human head. This example uses eight modular generators 118 and assorted moderator materials coupled with reflecting and shielding material (e.g. graphite 144). Secondary moderators (166 and 170) can be composed of one or more materials. There are moderator spacing blocks 128 in one embodiment composed of the same material High Density Polyethylene (HDPE), Ultra High Molecular Weight polyethylene (UHMW), or (PTFE (Teflon)) as the secondary moderators. Blocks of these materials fit in between the modular generators and are adjacent to each generator's pre-moderator. They act as mechanical spacers as well as moderator components. The outside of this region, between and behind the modular generators 118, is filled with either graphite or lead 144 to serve as a neutron reflector and shield. FIG. 14B also shows a side section view of the apparatus of FIG. 14A taken along a line through the top and bottom generators. There is additional moderator material in the front and behind the modular generators, extending a little above the pre-moderator. In our example, the cylindrical space 164 available for the patient's head is 52 cm deep and 30 cm in diameter. This space might be lined with 1-mm of cadmium 162 to shield against too large a thermal neutron dose to the patent's skin. Shield 162 is also shown in FIG. 14A. In other embodiments this space may be lined with 6LiF. The exemplary arrangement as illustrated in FIGS. 14A and B has a secondary moderator consisting of multiple layers of 40% Al and 60% AlF3 (166) and an additional moderating cylinder 170 of either 7LiF or D2O. These materials are shown to be concentric rings in FIG. 14A. Since 7LiF or D2O can be expensive, thicknesses were varied to obtain a desired neutron beam quality without over-using either 7LIF or D2O. In the example shown in FIGS. 14A and 14 B the thickness ratio between the two segments is altered, the total moderator thickness is 34 cm, and the sources are R=52.5 cm from the origin (center of the brain). The effect of doing this varying these materials is plotted graphically in FIG. 15. The reflector material graphite 144 is 30 cm thick in this example, the thickness of the Teflon 168, t, in front of the 2.5 MeV source is varied, and the portion 170 of the moderator is either 7LiF or D2O. As t changes, the thickness of the Al/AlF3 166 of the moderator changes, with all other dimensions remaining constant. The target is embedded in the Teflon 168, UHMW or HDPE. Sources are titanium targets 106 being bombarded by deuterium ion beams 5.0 cm in diameter. Each target is emitting 4×1010 neutron/sec. Eight modulator generators 118 emit 3.2×1011 n/s total emission. A concentration of 10B in the tumor and health tissue (e. g. skin) is known to be possible. 10B tumor concentration is assumed to be 50 ppm, while 10B in healthy tissue is 15 ppm. The relative biological effectiveness (RBE) for 10B in tumor is 2.7, and in healthy tissue is 1.3. Tumor and healthy tissue doses are calculated using the NRC and ICRP models for neutron RBE. The material 7LiF was the best performer and D2O was second best. An important main objective in these examples is to give a sufficient dose of neutrons to the cancer while minimizing the dose to the healthy tissue and not damaging it. FIG. 15 shows the performance for moderators with different values for tin cm and either 7LiF or D2O in the secondary moderator. The ordinate R is the ratio of tumor dose at the origin to healthy tissue skin dose, and the tumor dose at the center of the brain assumed to be the site of the cancer. As can be seen from FIG. 15, 7LiF outperforms D2O. The best performance is R=1.9 and a tumor dose in excess of 1.4 Sv/hr. A consequence of RBE is that a small percentage of fast neutrons is essential to obtain a high value for R; also, a reasonable number of epithermals is required to penetrate the target. Thus a combination of 7LiF and D2O may outperform either material alone. A Need for Small Animal Neutron Sources Development of boron delivery agents for BNCT is an ongoing and challenging task of high priority. A number of boron-10 containing delivery agents have been prepared for potential use in BNCT. With the development of new chemical synthetic techniques and increased knowledge of the biochemical requirements needed for an effective agent and their modes of delivery, a wide variety of new boron agents has emerged, but only two of these, oronophenylalanine (BPA) and sodium borocaptate (BSH) have been used clinically and have US FDA approval. Patient-derived xenograft (PDX) is created by transferring primary tumors from a patient into a mouse or small animal model. Tests of delivery and effectiveness of drugs to the cancer site can then be performed. In the prior art, only beamlines from nuclear reactors and linear accelerator structures can be used. A small laboratory neutron source, as in embodiments of this invention, is therefore valuable in the development and testing of new boron delivery drugs and their effectiveness in destroying the cancer site. As compared to a clinical delivery system, a smaller number of stand-alone generators such as generators 118 is needed for a delivery system for a small animal such as a mouse. The modular generators used have a slab wall angle of α=0 (see α defined in FIG. 12 A). The secondary moderator may be a separate container of heavy water (D2O). Since the small animal target is indeed small, the secondary moderator volume can be reduced, and the compact modular generators can be moved close to it permitting the modular generators to be closer to the animal target. Thus, the neutron flux at the cancer site is increased, and with proper selection of moderator material and size, will still be able to moderate the neutrons to IAEA standards. In addition, by moving closer, the number of generators can be reduced while still maintaining a high thermal neutron flux at the cancer site. In our example of the new art for a small animal source, we can use four modular generators 118 to emit enough thermal neutrons at the cancer site. We can use the modular generators of 12 A, B, C with the slab wall angle of α=0. This makes the pre-moderator 108 a rectangular cuboid (or “rectangular slab” of). FIG. 16A is a perspective view of a modular generator having such a rectangular pre-moderator 108, making it suitable for arrangements of four generators in a rectangular array, as shown in FIG. 16B. In FIG. 16B the four modular generators are arranged around a secondary moderator 112, which in one embodiment may be a container of heavy water or granulated moderator material. FIG. 16C is a cross section view of the arrangement of FIG. 16B, taken along section line 4 16C-16C of FIG. 16B. The elements previously annotated for modular generators are reused in FIGS. 16 A, B and C. FIG. 16D is an exploded view where the four generators 118 are moved back from the small heavy water moderator 112. Each generator 118 has a pre-moderator 108 with a fast neutron generator with a titanium target 106. A deuterium ion beam is generated by a plasma ion source 102 and accelerated in an acceleration chamber 100 to the titanium target 106, where the DD fusion reaction occurs releasing fast 2.5 MeV neutrons. This description is all common to the descriptions or other embodiments in the specification. The neutrons generated pass through a pre-moderator 108, where they are partially moderated to thermal neutron energies. They then pass into the moderator block 112 where they are further moderated, reducing the energy of fast neutrons to thermal neutron energies. The thermal neutrons then enter a cylindrical mouse chamber 114 where they enter the small animal 116. The pre-moderator is designed to slow the fast neutrons to thermal neutrons by scattering the fast neutrons via collisions with the hydrogen in the HDPE or UHMW plastics. The distance the 2.5 MeV neutrons have to traverse is approximately 3 to 5 cm, wherein approximately 50% of the neutrons lose enough of their energy to be classified as thermal neutrons. These neutrons, containing both thermal and fast neutron components, can then travel into the moderator box 112, where they are further moderated by collisions with deuterium atoms. Roughly speaking, the HDPE with its hydrogen-atoms moderates the neutrons to thermal energies over a short distance; the thermalized neutrons then penetrate the cylindrical chamber 114 wherein they place the small animal 116. The small animal model is used to test the delivery of boron to the cancer site. For the pre-moderator, high density polyethylene (HDPE) is optimum for producing the maximum flux of thermal neutrons. As in the case of the clinical generator, it is desired to produce a maximum thermal flux at the cancer site. A mouse is a small object, and penetration of thermal neutrons to the cancer site can easily be achieved. Moderation of the fast neutrons to thermal energies is desired with minimum production of gamma radiation, which is harmful to the healthy cells. As those skilled in the art will understand, hydrogen atoms are excellent at scattering fast neutrons, resulting in moderation of the neutrons to thermal energies in the shortest path length in the moderating material. Indeed, using 5-6 cm of high-density polyethylene (HDPE) or UHMW plastic results in moderation of about 50% of 2.5 MeV neutrons to thermal energies. Further moderation of the neutrons by longer distances in the HDPE results in more fast neutrons being converted to thermal energies. However, this results in reduction of the total flux (n/cm2) that is available since the neutrons are being emitted in a 4π solid angle. Hydrogen capture of neutrons produces high energy gamma radiation, which is destructive to both healthy and cancerous cells. Adding another moderator to further thermalize the neutrons is accomplished by the use of heavy water (D2O). The skilled person will understand that the embodiments described in this application are exemplary, and not limiting. Many variations may well fall within the scope of the invention, which is limited only by the scope of the following claims. |
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abstract | A process signal control and monitoring system, includes: a signal processing device which is installed on an outside of a nuclear reactor containment vessel, an internal electrical power source, an analog-digital conversion part, an internal communication part which transmits the digital signal to the signal processing device, an internal repeater, and an external repeater which transmits the received signal to a communication satellite. When electric power supply from the signal processing device is disconnected, the internal electrical power source supplies electric power which is charged in the rechargeable battery, to the analog-digital conversion part and the internal communication part; and the internal communication part judges whether communication with the signal processing device is continued or disconnected; and when the communication is judged to be continued, the internal communication part continues transmitting the digital signal to the signal processing device. |
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049833537 | description | Referring to FIG. 1 a sodium cooled nuclear reactor vessel V is illustrated. The particular vessel V shown is a prior art configuration. A steam generator M is illustrated operatively connected to the sodium reactor. The steam generator here shown is a preferred embodiment of a related art steam generator. This steam generator is not prior art. A complete description of this generator may be found in U.S. patent application Ser. No. 231,031 filed Aug. 11, 1988 and entitled Compact Intermediate Heat Transport System for Sodium Cooled Reactor, now U.S. Pat. No. 4,905,757, issued Mar. 6, 1990. Referring to the reactor V, a pool of sodium 14 is confined within an inner shroud vessel 16. Sodium pool 14 forms the so-called hot leg. Tracing the hot leg of the sodium cooled nuclear reactor, sodium from the sodium pool 14 passes upwardly from a core 12 where it receives heat. It thereafter passes downwardly through an intermediate heat exchanger H. In such passage it liberates its heat to the "cold leg" of the secondary loop. After the liberation of heat, the sodium of the primary loop then passes in its own "cold leg" to a bottom plenum 20. In bottom plenum 20 the sodium passes upwardly in a pumping leg at an annulus 22 into pump inlet 24 and through an electromagnetic pump P1. At electromagnetic pump P1, the sodium reverses at loop 26 passing through a discharge plenum 28 to the bottom of the core 12. At core 12 the sodium flows upwardly and to pool 14. The cycle endlessly repeats. It will be noted that sodium flow occurs within an inner shroud L. Shroud L provides an emergency heat outflow Such emergency heat outflow is not pertinent to this disclosure and will not be further discussed here. As is common in reactors, a control rod cavity 30 contains applicable control rods for the penetration into and out of the reactor to control the reaction Intermediate heat exchanger H interior of the sodium cooled reactor vessel constitutes the heat exchange interface between the primary and radioactive sodium loop and the secondary sodium loop. As here illustrated, lines 18, 20 provide for secondary sodium flow to and from the intermediate heat exchanger H. As here illustrated, the line 20 is a part of the cold sodium leg of the secondary loop. The line 18 is a part of the hot sodium leg of the secondary loop. Hot sodium flows in outer concentric pipe 18 into the steam generator M. Generator M constitutes a generally cylindrical vessel with dome closures at both ends and having an outer vessel 60 and an inner and concentric vessel 62. The interstitial volume between the outer vessel 60 and the inner concentric vessel 62 is filled with helically coiled tubes. These tubes begin at tube sheets placed within lower water inlets 71, 74. The tubes extend upwardly into the interstices between the outer vessel 60 and the inner vessel 62. Specifically, and in the area 78, the tubes coil helically about the inner vessel 62. In such helical coiling, the tubes coil until they reach the upper portion 78 of the steam generator M. At upper portion 78, the tubes pass directly vertically upward to tube terminating tube sheets within steam outlets 81, 84. The steam is generated by the heat transferred from the hot sodium during the upward passage of water through the helically coiled tubes. The hot leg of the secondary sodium loop continues at inlet pipe 40. Sodium counterflows the water in the helically coiled tubes 78. This counterflow includes passage from the inlet at 40 down to the plenum 64. At plenum 64, upward sodium flow occurs in two separate paths. First, a single electromagnetic pump Q' is located. Pump Q' takes suction at 201 and discharges high pressure, sodium at 200. The discharged high pressure, sodium passes into the inlet 210 of a jet pump located inside of the interior cylindrical vessel 62 and supported by struts 240 In the second flow path, sodium flows interior of the inner vessel 62 outside of the electromagnetic pump. Specifically, and as indicated at arrow 250, sodium flows in an annulus exterior of the electromagnetic pump and passes into the mixer section 210 of the jet pump. The sodium then exists at a diffuser 220 into an outlet 230. At outlet 230 the sodium is pumped to and towards the heat exchanger H. It will be seen at the bottom of the steam generator that there is provided a diaphragm D mounted to a protruding nozzle 270. Diaphragm D is designed to rupture in the case of a sodium water reaction. When the diaphragm D ruptures, sodium empties from the steam generator vessel. Having set forth the prior art sodium reactor vessel V and the related art steam generator M, the casualty scenario against which this invention guards may now be set forth. It is assumed for purposes of the discovered scenario that a tube rupture has occurred in the worse possible location. Specifically, such a location is shown at 300. It is further assumed that more than one tube is effected by the rupture and the pressure generated by the chemical reaction breaks the rupture diaphragm. Viewing FIG. 1, it can be seen that the sodium, hydrogen, steam and other compounds from the violent sodium water reaction at 300 have to pass along the entire length of intact tubes within the coiled helical tubes 78 to the rupture diaphragm. After such passage, the gases will find their way into plenum 64 and out diaphragm D at protruding cylindrical nozzle 270 at the bottom of the steam generator. It will be remembered that high pressure steam in lines 90 and high pressure feedwater in lines 91 is assumed to be present. This high pressure steam from the turbine side of the plant and feedwater from supply steam is presumed to flow to the site of the reaction at 300. Accordingly, region 300, the site of the tube breakage, will be presumed to be a high pressure violent reaction continuously supplied with the necessary sodium reactive steam and water to keep the reaction sufficiently long (terminated in the steam generator by Na expulsion thru the rupture disk) to cause a large number of tube ruptures at the site 300. This being the case, the present casualty scenario presumes that the continuing steam/water flow and associated pressure drop within the tube bundle will force the Na/steam interface along conduits 18, 20 and back into the intermediate heat exchanger H. It will be realized that as the steam sodium interface penetrates the specific conduits 18 and 20, the conduits will, in all likelihood, propagate the sodium water reaction into the main reactor vessel. Remembering that the sodium interior of the vessel is radioactive, complication of the disclosed casualty by penetration of the steam into the radioactive vessel is to be avoided. This being the case, the improvement of this invention can now be set forth. Referring to FIG. 2, a steam generator M having an outer cylindrical vessel 60 and an inner cylindrical vessel 62 is illustrated. Between inner cylindrical vessel 62 and the outer vessel 60 there is placed an intermediate cylindrical vessel 63. Intermediate vessel cylindrical 63 opens to the plenum 64 at the bottom. Likewise, intermediate cylindrical vessel 63 opens at the top to the cover gas region C. In the view illustrated in FIG. 2 normal reactor operation is assumed. It is instructive to understand this normal reactor operation so that a serendipitous advantage of this invention can be understood. Sodium typically flows in from the reactor along leg 18 and is distributed at a manifold 170 at the top of the reactor. The sodium in the hot leg flows downwardly over the helical tubes 78 down into plenum 64. At the plenum 64 the sodium flows inwardly to the inside of the interior cylindrical vessel 62. At this point, pump Q' acting as an electromagnetic pump, pumps a high pressure, low volume, flow of sodium into a jet pump inlet 210. The sodium discharged from the electromagnetic pump entrains sodium passing about the outside surface of the pump into the mixing section of the jet pump 210. The sodium passes to a diffuser 220 and outwardly on the cold leg 20. It will be appreciated that plenum 64 is the low pressure region of the secondary sodium loop. The sodium in the interstitial area between the inner cylindrical vessel 62 and the intermediate cylindrical vessel 63 is supported in its static head from the relatively low pressure plenum 64. Plenum 64 has a relative low pressure because it constitutes the suction side of the pump Q'. Consequently, it has a sodium/cover gas interface 80 adjacent the bottom of the interstitial volume between the inner cylindrical vessel 62 and the intermediate cylinder 63. It can be seen that the cover gas C penetrates downwardly almost the full length of the intermediate cylindrical vessel 63. There is thus placed between the inner cylindrical vessel 62 and its cold leg of sodium and the outer vessel 60 and its contained hot leg of sodium, a region of cover gas C. Insulation by the region of cover gas C occurs not unlike that insulation common in a Dewar flask. Stated in other terms, the intermediate cylinder 63 prevents heat being shunted directly from the hot leg to the cold leg of the steam generator M. Having set forth this serendipitous characteristic, operation of the steam generator in the casualty scenarios herein set forth can be understood with respect to FIG. 3. Assuming that a casualty has occurred in an area 300, the diaphragm D on cylindrical nozzle 270 at plenum 64 immediately ruptures. Liquid sodium from the secondary loop immediately drains to a sodium dissipation system including a holding tank and stack. These conventional prior art systems are not shown. Regarding the sodium in the interstitial volume between the inner cylindrical vessel 62 and the intermediate vessel 63, sodium likewise immediately drains. This draining of sodium opens a gas free path from the top of the steam generator C directly to the plenum 64. This can be seen to be almost direct from the site of the violent sodium water reaction 300. This may be easily understood. Assuming that a chemical reaction has occurred at 300 and steam is continuously being supplied, two flow paths will be present. First, steam can discharge from the site of the reaction down through the remaining intact tube 78 and out the diaphragm D. Since the remaining intact tubes constitute a considerable flow barrier, especially where the tube rupture is in the upper portion of the tube coils, this route for the outgassing of the components of the violent reaction will have only a minority of the total flow. An additional flow path is defined between the inner vessel 62 and the outer vessel 63. Specifically with all sodium expelled, gas can pass upwardly from the site of the 25 reaction into the now vacated cover gas region C' and in the top of the intermediate cylindrical vessel 63. From the intermediate vessel 63, a direct and free nonencumbered flow path out the diaphragm D is defined. Consequently, hot leg inlet 18 and cold leg outlet is 20 does not experience a large pressure differential. Specifically, steam/water from the site 300 cannot penetrate along the length of conduit 18 to effect the continuance of the casualty to and towards the reactor. It should be mentioned that because of the intermediate cylinder 63, the overall diameter of the steam generator vessel 60 is slightly increased. However that may be, the increase is not substantial. For example, whereas a prior art steam generator illustrated in FIG. 1 as a diameter of 8 feet, the disclosed generator with the intermediate vessel has a diameter of 9 feet. It will be understood that this invention can be operative in those types of steam generators which do not include a central contained pump. Such a steam generator is illustrated in FIG. 4 Referring to FIG. 4, a steam generator M' is illustrated having an outer vessel 60 and a single interior cylindrical vessel 63. Vessel 63 opens to a plenum 64 at the bottom and opens to the cover gas region C at the top. As before, helically coiled tubes conventionally run between feedwater inlets 71, 74 at the bottom and steam outlets 81, 84 at the top. In most steam generator constructions, it is not possible to helically coil the tubes 78 to occupy the entire inner diameter. Consequently, and in the prior art, an inner cylindrical vessel 63 has normally been a vacuous and closed area. Sodium is conventionally withdrawn from plenum 64 in the cold leg and passed to a relief nozzle (the relief nozzle not being shown in the view of FIG. 4). The reader can understand that the installed conduit 63 without the inner cylinder 62 functions precisely analogous to that illustrated in FIGS. 2 and 3. Specifically, and during normal operation (as shown in FIG. 2) the sodium level in the central duct will be at an elevation supported by the low pressure in the plenum 64. Upon a casualty occurring at the top of the coiled tubes, sodium will empty, and the gas and sodium from the site of the violent reaction will pass interiorly of the central cylinder 63 and out the bottom of the vessel. The reader will likewise appreciate that varying constructions may be used. |
044617224 | summary | It is the endeavor of the technical experts in the art to create the possibility of safely handling waste materials, which have no further industrial use, during their transport or storage so as not to endanger the environment. One of the requirements to accomplish this, for instance during the transport or storage of aqueous solutions which contain radioactive materials, consists in treating these solutions prior to their transport or storage so as to convert them into solid end products. However, in this connection it is not sufficient to confine or seal the aqueous solution in permanent containers. A heretofore known method of solidifying aqueous solutions which contain radioactive materials consists in adding the solutions which result from chemical separations, activation anaylses, extractions, decontamination operations, or also from recovery of fuels, to a mixture of cement and vermiculite, whereby a solidification of the aqueous solution from the reaction with the cement is obtained. However, a drawback of this known method consists in that a troublesome development of gas and heat during the solidification of aqueous, acidic solutions occurs, leading to long delays and, therefore, making the method uneconomical. In addition thereto, during solidification of strongly acidic aqueous solutions, a safe accomplishment of the method can no longer be assured for the operating personnel since, because of the great heat development, bubbling-up and spattering of the solution cannot be avoided. Furthermore, there also exists the possibility of contaminating the environment. A further drawback consists in that the vermiculite contained in the mixture, because of its light weight, is partially carried on the surface of the aqueous solution, resulting in a non-homogeneous end product which does not meet the requirements for the solidification of the aqueous solution. It is an object of the present invention to provide a method of solidifying aqueous solutions containing radioactive or toxic materials, which makes it possible to produce an end product which contains the waste materials in a homogeneous distribution and can be produced in such a manner as to be safe to the operating personnel. It is a further object of the present invention to provide a method as set forth in the preceding paragraph which can be carried out even if the aqueous solution to be solidified is strongly acidic or alkaline, without having to make allowance for, or put up with, long delays. Yet another object of the present invention consists in that the substances necessary for carrying out the method should be as inexpensive as possible. With these and other objects and advantages in mind, the method according to the present invention is characterized primarily in the aqueous solution, which contains one of the mineral acids, such as HF, H.sub.2 SO.sub.4, HClO.sub.4, HCl or HNO.sub.3, or one of the alkalies, such as KOH, NaOH, NH.sub.3 or Ca(OH).sub.2, up to 40% by weight, or water soluble organic compounds up to 50% by weight, is mixed with a porous, solid substance having a carbonate content of less than 1%, the substance comprising a ceramic material, pumice, or the like having a granulation from about 2 mm up to an average diameter of about 20 mm. The aqueous solution is also mixed with gypsum having 4.7 to 6.6% water of crystallization and a carbonate content of less than 1%. The mixture ratio of gypsum to porous, solid substance is from 1 to about 0.5 to 3, and the mixture ratio of the aggregate of gypsum and porous, solid substance to aqueous solution is about 0.7 to 1.3 kg to 500 ml. No disturbing gas or heat development occurs while carrying out the method according to the present invention. Since by using the porous, solid substance, which may be comprised, for example, of ceramic tile chippings having a specific weight between 1.0 and 1.4 kg/dm.sup.3, a good inter-mixture of the gypsum with the waste materials contained in the aqueous solution is realized, the waste materials are correspondingly homogeneously distributed in the solid end product. During mixing, the components of the mixture of gypsum, solid substance, and aqueous solution are expediently mechanically agitated by means of a stirring apparatus, agitator, or the like, and the requisite amounts are added in a sequence adapted or proportional to the requisites of the mixture. In this connection it may be expedient by batches or continuously to add the quantities of gypsum, solid substance, and aqueous solution. In the event that also non-aqueous organic compounds are used, alcohol and water are added to these compounds in such an amount that the mixture contains about 20% non-aqueous organic compound, the thus formed mixture of non-aqueous organic compound, alcohol, and water is then intermixed with an appropriate amount of gypsum and solid substance. A particularly advantageous specific embodiment of the method according to the present invention consists in first intermixing the gypsum and porous, solid substance and subsequently adding the aqueous solution to the thus formed mixture. This makes it possible to produce a homogeneous end product without necessitating a mechanical agitation of the components which are to be intermixed. Since, in addition, the gypsum and solid substance mixture may be stored for several months in a closed container without becoming unusable, the solidification of aqueous solution is a very simple manner is possible. To do so, it is merely necessary to pour the mixture of gypsum and solid substance into the container intended for the final storage and then to add the aqueous solution. An alternative solution to the previously stated objects as taught by the method according to the present invention is characterized primarily or first in the substance which promotes the intermixing of binding agent and aqueous solution be so formed that gypsum having a crystal content of about 4.7 to 6.6% and a carbonate content of less than 1% be mixed together with water glass having a specific gravity in the range of from 1.2 to 1.8 kg/dm.sup.3 ; the mixing proportion is about 1 kg gypsum to 100 to 500 ml water glass. To the thus formed, partially granular, partially pulverous mixture there is added the aqueous solution which contains one of the mineral acids, such as HF, H.sub.2 SO.sub.4, HCIO.sub.4, HCl, or HNO.sub.3, or one of the alkalies, such as KOH, NaOH, NH.sub.3, Ca(OH).sub.2, up to 40% by weight, or water soluble organic compounds up to a content of 50% by weight; the mixing proportion is about 1 kg of the gypsum and water glass mixture to 500 ml aqueous solution. Sodium as well as potassium water glass (Me.sub.2 SiO.sub.3, Me.sub.2 SiO.sub.4, Me.sub.2 SiO.sub.5) may be used. With this alternative solution also no disturbing gas or heat develops if strongly acidic or alkaline aqueous solutions are being solidified. When intermixing gypsum and water glass, it is expedient to add water glass by batches while stirring the added water glass into the gypsum. The thus formed, partially granular, partially pulverous mixture, which may be stored for about a week, for the solidification of an aqueous solution is placed in a container intended for the final storage, and the aqueous solution is added. In this connection the advantage is obtained that no mechanical agitation of the mixture is required, so that also by this alternative method according to the present invention a simple and safe solidification of the aqueous solution may be carried out. |
056129831 | claims | 1. A device for filtering water to at least one emergency cooling system in a nuclear power plant of the type comprising a reactor arranged in a containment which substantially consists of an upright, suitably cylindrical container whose bottom part forms a pool for collecting water formed by condensation of steam present in the containment, the condensation pool including a number of back-flushable strainers (3) serving to filter water which is taken from the pool and, if required, is supplied to nozzles in the emergency cooling system in order to cool the reactor core in the event of an inadmissible temperature rise therein, each strainer having the shape of a housing with at least one, suitably cylindrical, apertured strainer wall (10) through which the water can flow from the outside and into the housing, and being connected, by a first conduit (4) passing through the container wall (1), to a suction pump disposed outside the container wall, as well as connected to a second conduit (7) for supplying wash water to the interior of the housing in order, if required, to flush the strainer wall (10) by flowing the wash water through it from the inside and out, thereby removing filtrate deposited on the outside of the strainer wall, characterised in that a number of secondary strainers (24), each consisting of an elongate, apertured tube which is substantially vertically mounted and has a diameter or maximum cross-sectional dimension from about 200 mm to about 400 mm and a length dimension at least five times greater than the diameter dimension, are connected either directly or indirectly by a third conduit (23) to the first conduit (4) connected to the suction pump. 2. A device as set forth in claim 1, characterised in that the secondary strainer tubes (24) are mounted on a vertical upright, preferably in the form of a column (2) arranged in the containment and forming part of the load-bearing structure thereof. 3. A device as set forth in claim 2, characterised in that the structure is achieved by means of a number of clamp sets which are vertically spaced apart along the upright (2) and which each comprise a main clamp (25) consisting of two first part-circular hoop elements (25', 25") which enclose the upright (2) and are interconnected by bolted joints (26, 26'), and that at least one of the two first hoop elements (25', 25") has a number of radially projecting support means (27) which at a free end support one of two second particular hoop elements (28', 28") which together enclose an individual strainer tube (24) and are interconnected by bolted joints (29, 29') or the like. 4. A device as set forth in claim 1, characterised in that the strainer wall (10) of the back-flushable strainer (3) includes means (17) for dividing a fibre mat or layer built up on the outside of the strainer wall into several part sections. 5. A device as set forth in claim 4, characterised in that the means (17) consist of a number of longitudinal, peripherally spaced-apart and radially projecting wings or wing-like elements (17). 6. A device as set forth in claim 4, characterised in that the second conduit consists of a tube (7) which, in the area of an opening located concentrically in relation to the strainer wall (10), includes rotation-generating means (18) comprising a set of curved blades (20) extending between the tube and a substantially conical body located centrally therein, whereby water passing through the annular gap between the tube and said body is caused to rotate or circle so as to be pressed out against the strainer wall (10). 7. A device as set forth in claim 1, characterised in that the housing of the back-flushable strainer (3) includes an upper end and a lower end, said strainer being closed at said upper end the first conduit (4) being connected to the lower end, and the wash-water conduit (7) being concentrically inserted in a portion (4') of the first conduit (4). 8. A device as set forth in claim 1, characterised in that the third conduit (23) is connected to the first conduit (4) at a point (30) between the back-flushable strainer (3) and the suction pump. 9. A device as set forth in claim 2, characterized in that the strainer wall (10) of the back-flushable strainer (3) has means (17) for dividing a fibre mat or layer built up on the outside of the strainer wall into several part sections which separately are more easily released than a continuous circumferential fibre mat. 10. A device as set forth in claim 3, characterized in that the strainer wall (10) of the back-flushable strainer (3) has means (17) for dividing a fibre mat or layer built up on the outside of the strainer wall into several part sections which separately are more easily released than a continuous circumferential fibre mat. 11. A device as set forth in claim 5, characterized in that the second conduit consists of a tube (7) which, in the area of an opening located concentrically in relation to the strainer wall (10), is provided with a rotation-generating means (18) in the form of a set of curved blades (20) extending between the two and a substantially conical body located centrally therein, so that water passing through the annular gap between the tube and said body is cause to rotate or circle so as to be pressed out against the strainer wall (10). 12. A device as set forth in claim 1 wherein said maximum cross-sectional dimension of said secondary strainer is between about 250 mm and 350 mm. 13. A device as set forth in claim 1, characterized in that said length dimension is at least ten times greater than the diameter dimension. 14. A device for filtering water to at least one emergency cooling system in a nuclear power plant of the type comprising a reactor arranged in a containment which substantially consists of an upright, suitably cylindrical container whose bottom part forms a pool for collecting water formed by condensation of steam present in the containment, the condensation pool including a number of back-flushable containers (3) searing to filter water which is taken from the pool and, if required, is supplied to nozzles in the emergency cooling system in order to cool the reactor core in the event of an inadmissible temperature rise therein, each strainer having a shape of a housing with at least one, suitably cylindrical, apertured strainer wall (10) through which the water can flow from the outside and into the housing, and being connected, by a first conduit (4) passing through the container wall (1), to a suction pump disposed outside the container wall, as well as connected to a second conduit (7) for supplying wash water to the interior of the housing in order, if required, to flush the strainer wall (10) by flowing the wash water through it from the inside and out, thereby removing filtrate deposited on the outside of the strainer wall, characterized in that a number of secondary strainers (24), each consisting of an elongate, apertured tube having a diameter or maximum cross-sectional dimension from about 200 mm to about 400 mm and a length dimension at least five times greater than the diameter dimension, are connected either directly or indirectly by a third conduit (23) to the first conduit (4) connected to the suction pump. |
055880312 | description | DETAILED DESCRIPTION OF THE INVENTION In the following description, like reference characters designate like or corresponding parts through-out the several views of the drawings. Also, in the following description, it is to be understood that such terms as "forward," "left," "right," "upwardly," "downwardly," and words of similar import are words of convenience and are not to be construed as limiting terms. Referring to FIG. 1, a reactor vessel 10 is shown for converting water to steam. The reactor vessel includes an upright pressure housing 20 having a cylindrical wall 25 which is closed at its lower portion by a dish-shaped bottom head 30 having downwardly extending feet 40 which rest on a foundation (not shown), typically a floor. The upper end of the pressure housing 20 is closed by a removable dome-shaped top head 50 secured to the pressure housing 20 by a plurality of nuts 60 and stud bolts 70 (only one of each is shown). The top head 50 is sealed by a gasket 100 to the upper end of the pressure housing 20 for forming a secondary seal therebetween. Steam dryer panels 110, which may be of a conventional type, are mounted in the upper end of the vessel 10 for drying the steam as it passes therethrough. Water is maintained in the lower portion 115 of the vessel 10 for providing a medium which will convert to steam. Steam separators 120, which also may be of conventional construction, are mounted in the vessel 10 just below the steam dryer panels 110 for separating the properly dried steam from the steam containing water in excess of a predetermined amount. A plurality of separator vapor tubes 130 extend down respectively from each steam separator 120 and is sealed through a steam plenum cap 140 of a cylindrical shroud 150 disposed coaxially within the pressure housing 20 to leave an upright space or downcomer annulus 160 between the shroud 150 and the housing 20. The shroud 150 has a generally tapered configuration with its upper portion having a greater diameter than its middle portion, and its middle portion having a slightly greater diameter than its lower portion. Feed water is supplied to the reactor vessel 10 through four feed water sparger nozzles 170 (only one sparger nozzle is shown) located at equal intervals in a horizontal plane. A reactor core fuel assembly 180 is made up of a plurality of elongated vertical fuel assemblies 190 which are arranged in groups of four. The lower end of each fuel assembly 190 in each group rests on a vertical respective control rod guide tube 200 sealed at its upper end through a horizontal bottom grid plate 210 mounted across the bottom of the shroud 150. Each guide tube 200 extends down below the bottom grid plate 210, and a separator control rod 215 is mounted in each control rod guide tube 200 to slide longitudinally up and down between the four adjacent elongated vertical fuel assemblies 190 resting on the grid tube 200. Thimble tubes 225 are positioned in the lower portion of the reactor vessel 10 for receiving control rods (not shown) which control the nuclear reaction in the reactor vessel. Water flows upwardly through the fuel assemblies 190 where water changes to steam, and then passes as a steam-water mixture out of the vapor tubes 130 and through the steam separators 120. Water separated from the steam in the separators 120 is returned to the downcomer annulus 160. Steam passes the steam drier panels 110, and leaves the vessel 10 through a steam outlet 220 to pass through a conventional steam turbine and condenser (both of which are not shown), as is well known in the art. Condensed steam is returned from the condenser to the feed water sparger nozzles 170 by a conventional pump (not shown). The lower end of the shroud 150 is welded to the upper end of a cylindrical shroud support skirt 230, the lower end of which is welded to an annular ring 240 formed integrally with the bottom head 30 of the vessel 10. An annular shaped support structure 245 extends around the core shroud 150 and between the core shroud 150 and the pressure housing 20 for assisting in supporting the core shroud 150. As will be discussed later in detail, a plurality of reinforcing devices 247 (only one of which is shown) are attached to the core shroud 150 in a spaced apart relationship with each other for reinforcing the core shroud 150 in the event of cracking in the core shroud 150. A core inlet plenum chamber 250 is formed within the shroud support skirt 230 and between the bottom grid plate 210 and the bottom head 30 of the vessel 10. Referring to FIGS. 2 and 3, one of the plurality of reinforcing devices 247 is shown in detail. The device 247 includes a vertically oriented beam 260 positioned against the core shroud 150 for reinforcing the shroud 150 in the event of stress corrosion cracking. Such cracking is likely, if at all, to occur where an upper portion of the shroud 150 joins its middle portion (generally designated by 270), and its middle portion joins its lower portion (generally designated by 280). The beam 260, which has a generally U-shaped horizontal cross section, includes a base section 290 having two lip portions 300a and 300b each respectively projecting outwardly from the outer portions of the base section 290. At the bottom of the beam 260, a base plate 310 is attached to the beam 260 for allowing the beam 260 to rest on the annular ring 245. The base plate 310 is attached to the annular ring 245 by a bolt 320 for rigidly attaching the beam 260 to the annular ring 245. The bolt 320 further includes a washer 330 for ensuring a tight fit to the annular ring 245. To enable the beam 260 to be supported at its top portion, a notch 340 is made into the shroud 150 by any suitable means, such as by an electro-machine discharge (EMD) which is well known in the art. A bolt 350 rests in the notch 340 with its bolt head 360 positioned in the notch 340 and with its shaft 370 projecting through the base section 290 of the beam 260. A nut 400 and washer 410 are placed on the shaft 370 for firmly positioning the beam 260 and shroud 150 into their respective operating positions. Once the beam 260 is placed in its vertical position, three radial supports 420 are positioned in a spaced apart relationship with each other along the beam 260 and between the beam 260 and the reactor wall 25 for transmitting the forces absorbed by the beam 260 from the shroud 150 to the reactor wall 25. The supports 420 also function to uniformly distribute the absorbed forces of the beam 260 along the beam 260. A portion 430 is attached between the base section 290 and the shroud 150 on the two lower radial supports 420 for, in addition to the attachment of the beam 260 to the shroud 150 at its top and bottom portions, transmitting forces from the shroud 150 to the beam 260. The portion 430 may be made of 316 stainless steel. Referring to FIG. 4A, a portion of the radial support 420, which is attached at the extreme bottom of the beam 260, is shown in detail. The beam 260 includes two generally rectangular shaped support shelves 440a and 440b respectively welded against each lip portion 300a and 300b of the beam 260 for supporting a mating portion 450 (see FIG. 5) of the radial support 420. As can be seen in FIG. 4B, the support shelves 441a and 441b at the middle portion of the beam 260 are shorter in length than the support shelves 440a and 440b at the bottom portion. The reason for this will be described later in detail. Similarly, as seen in FIG. 4C, two support shelves 442a and 442b positioned at the top of the beam 260 are shorter in length than the other support shelves 441a, 441b, 442a and 442b positioned below it. Each support shelf includes a hole 445 therein for receiving a bolt 460 (see FIG. 5) for attaching all shelves to its mating portion 450 (See FIG. 5). Referring to FIG. 5, the mating portion 450 of the radial support 420 for the bottom location is shown. The mating portion 450 includes a generally tapered top portion 470 having a notched out portion 480a and 480b at its two corners. The top portion 470 includes two holes 490 therein which respectively align with the mating holes 445 in the support shelves 440a and 440b for each receiving a bolt 460 when so aligned. The lower portion of the mated portion 450 also includes a generally tapered body 500 which also includes two notched out portions 510a and 510b which conform generally to the shape of the support shelves 440a and 440b. To install the mated portion 450 to the support shelves 440a and 440b at the lower portion of the beam, the mated portion 450 is placed in the interior of the beam 260 (see FIG. 3) with the notches 480a and 510a, which are positioned in one corner of the mated portion 450, positioned directly over the support shelf 440a which is positioned in the same corner of the beam 260. The other notches 480b and 510b on the opposite corner of the mated portion 450 are also so aligned with the other support shelf 440b. The mated portion 450 is slid down the interior of the beam 260 until the top portion 470 contacts the lower support shelves 440a and 440b. When in this position, the holes 450 and 490 are in registry with each other so that the bolts 460 can be placed therein for rigidly attaching the support shelves 440a and 440b and its mated portion 450. An end portion 520 is attached to the mating portion 450 for resting against the wall 25 of the reactor vessel 10 which, in turn, provides the mechanical communication between the wall 25 and the mated portion 450. Although only the mated portion 450 at the extreme lower portion was described, the mated portion 450 for the middle portion is the same except that the length (L) is longer so that it mates to its mating support shelves 441a and 441b; while at the same time, it has to be short enough so that it passes over the extreme top support shelves 442a and 442b when it is slid down the beam 260. It will also be apparent to those skilled in the art that all the notches 510a and 510b in the mated portion 450 at the middle and extreme top portion of the beam 260 are not quite as large as the notches in the mating portion 450 at the extreme bottom so that they will conform to their support shelves 441a and 441b. The mated portion 450 for the extreme top portion of the beam 260 is the same as the extreme bottom mated portion 450 except that the top portion 470 conforms to the entire shape of the interior of the beam 260, i.e., there are no notches in the top portion 470. It is thought that 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 without departing from the spirit and scope of the invention or sacrificing all of its material advantages, the form hereinabove described being merely a preferred or exemplary embodiment thereof. |
044118619 | summary | The invention of the instant application relates to a method for protecting the casing or cladding tubes of nuclear reactor fuel rods, which are formed of zirconium alloy, against the attack of radioactive fission products, such as iodine, especially. The fuel rods, especially of light-water reactors, are formed of a cladding or casing tube to which end caps have been welded at both ends thereof, and a charge of generally pellet-shaped nuclear fuel, such as uranium dioxide or a mixture of uranium dioxide and plutonium dioxide, for example. The wall thickness of the cladding tubes is about 0.3 to 0.7 mm. The operating temperature in the nuclear reactor is in the order of 400.degree. C., but this temperature can be exceeded for short periods of time in the event of accidents. Operating experience with such nuclear reactor fuel rods over many intervening years has shown that presently employed fuel rod constructions have a high degree of reliability. Occurrence of damage to nuclear reactor fuel rods has been rare and has required in most cases no immediate interruption of the operation of the respective nuclear reactor. This experience prevailed especially for such nuclear reactors which were operated with uniform load. It was found, however, that more fuel rod defects occurred when nuclear reactors had to be operated with varying load. This varying load resulted in varying temperatures of the nuclear reactor fuel rods, which were consequently also stressed cyclically, in addition, by thermal expansion. It was accordingly found that stress-crack corrosion set in at the inner wall surface of the cladding tubes, radioactive fission products, especially iodine, being found to be the corroding media. The problem therefore arose of finding ways and means to prevent such stress-crack corrosion to the greatest extent possible and, thereby, to make the fuel rods as reliable under varying-load operation as under constant load. It is accordingly an object of the invention to provide a method of protecting the casing or cladding tubes of nuclear reactor fuel rods from such stress-crack corrosion and for, thereby, rendering the fuel rods equally reliable under varying-load as well as constant-load operation. With the foregoing and other objects in view, there is provided, in accordance with the invention, a method of protecting a zirconium-alloy cladding tube of a nuclear-reactor fuel rod against attack by radioactive fission products, such as iodine especially, which comprises applying an internal pressure to the cladding tube at a temperature of 300.degree. to 500.degree. C. so as to deform the cladding tube, depending upon the geometric dimensioning thereof, in the elastic range and up to nearly the yield point thereof and, while this condition exists, reacting a medium previously introduced into the interior of the cladding tube with the inner surface of the cladding tube to form a protective layer. At the end of the empirically determinable time for the formation of the protective layer, the temperature and the pressure are lowered again so that the cladding tube resumes virtually the original dimensions thereof. The protective layer produced during this treatment is thereby compressed and is under great internal compressive stress. This stress does not also disappear during operation of the reactor because in the course thereof, a similar cladding-tube deformation cannot occur due to the externally applied high pressure of the coolant. It has now been found that cladding tubes of nuclear fuel rods treated in this manner have become virtually insensitive to stress-crack corrosion. This also corresponds to experience obtained in other fields of engineering with work pieces which were provided with an internal compressive stress zone as protection against stress-crack corrosion. In this connection, reference may be had to heat exchanger tubes which are provided on the outside thereof with a compressive stress layer by rolling or blasting with glass beads. In accordance with another measure of the method of the invention, the cladding tube is closed at both ends thereof, and the introduced medium serves simultaneously for applying the internal pressure thereto. In accordance with a further measure of the method invention, a given quantity of water serves as the introduced medium, the vapor thereof forming a protective layer of zirconium dioxide on the inner surface of the cladding tube. In accordance with an added measure of the method invention, a given quantity of H.sub.2 O.sub.2 serves as the introduced medium, the vapor thereof and liberated atomic oxygen forming a protective layer of zirconium dioxide on the inner surface of the cladding tube. In accordance with an additional measure of the invention, the method includes welding a plug beforehand to one open end of an open-ended tube so as to seal the respective open end thereof, thereafter introducing the given quantity of medium into the tube thus sealed at the one end thereof, temporarily closing off the other open end of the tube, then heating the tube to vaporize the medium that has been introduced therein and so as to cause it to react with the inner surface of the cladding tube, opening the temporarily closed other end of the tube after an empirically determined time period during which the protective layer is formed, and cooling the cladding tube. In accordance with yet another measure of the method invention, the medium is formed at least partly of a respective substance for increasing the internal pressure in the cladding tube and for forming the protective layer. In accordance with yet a further measure of the method invention, the substances are gaseous. In accordance with an alternate measure of the method invention, the substances are gas-forming. In accordance with a concomitant feature of the invention, the method includes filling the cladding tube sealed at the respective one end thereof with nuclear-fuel pellets, and sealing the other end of the cladding tube to form the fuel rod. Other features which are considered as characteristic for the invention are set forth in the appended claims. Although the invention is illustrated and described herein as embodied in a method for protecting the casing tubes of nuclear reactor fuel rods, it is nevertheless not intended to be limited to the details shown, since various modifications may be made therein without departing from the spirit of the invention and within the scope and range of equivalents of the claims. |
abstract | A compound x-ray lens and method of fabricating these lenses are disclosed. These compound lenses use multiple zone plate stacking to achieve a pitch frequency increase for the resulting combined zone plate. The compound equivalent zone plate includes a first zone plate having an initial pitch frequency stacked onto a second zone plate to form an equivalent compound zone plate. The equivalent zone plate has a pitch frequency that is at least twice the initial pitch frequency. Also, in one example, the equivalent zone plate has a mark-to-space ratio of 1:1. |
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summary | ||
abstract | An antenna system comprises a first end-fire antenna element and a second end-fire antenna element facing each other in a planar arrangement, the antenna elements being configured such as to cause destructive interference between individual end-fire radiations of the elements, while maintaining constructive interference generally perpendicular to the planar arrangement. |
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055815880 | summary | FIELD OF THE INVENTION This invention relates to reducing the corrosion potential of components exposed to high-temperature water. This invention relates to a method for reducing the corrosion potential of components exposed to high-temperature water by the use of protective, catalytic, insulating coatings. This invention is particularly related to the use of insulating coatings that are doped with noble metals and that contain restricted mass transport crevices that lower the corrosion potential of a coated metal component when the coating is in contact with high temperature water. BACKGROUND OF THE INVENTION As used herein, the term "high-temperature water" means water having a temperature of about 100.degree. C. or greater, steam, or the condensate thereof. High-temperature water is found in a variety of known apparatus, such as water deaerators, nuclear reactors, and steam-driven power plants. High temperature water may have elevated concentration of oxidizing species such as hydrogen peroxide and oxygen. Nuclear reactors are used in electric power generation, research and propulsion. A typical nuclear reactor comprises a reactor pressure vessel contains the reactor coolant, i.e. high temperature water, which removes heat from the nuclear core. Respective piping circuits carry heated water or steam to the steam generators or turbines and carry circulated water or feedwater back to the vessel. Operating pressures and temperatures for the reactor pressure vessel are about 7 MPa and 288.degree. C. for a boiling water reactor (BWR), and about 15 MPa and 320.degree. C. for a pressurized water reactor (PWR). Much of a nuclear reactor is fabricated from metal components comprising various materials. The materials used in both BWRs and PWRs must withstand various loading, environmental and radiation conditions, including exposure to high temperature water. Some of the materials exposed to high-temperature water include carbon steel, alloy steel, stainless steel and other iron-base alloys, as well as nickel-base, cobalt-base and zirconium-base alloys. Despite careful selection and treatment of these materials for use in water reactors, corrosion occurs on these materials when exposed to the high-temperature water. Such corrosion contributes to a variety of problems, for example, stress corrosion cracking, crevice corrosion, erosion corrosion, sticking of pressure relief valves and buildup of the gamma radiation-emitting Co-60 isotope. Stress corrosion cracking (SCC) is a known phenomenon occurring in metal reactor components, such as structural members, piping, fasteners and welds that are exposed to high-temperature water. As used herein, SCC refers to cracking propagated by static or dynamic tensile stressing in combination with corrosion at the crack tip. The metal components of a reactor are subject to a variety of stresses associated with, for example, differences in thermal expansion, the operating pressure needed for the containment of the reactor cooling water, and other sources such as residual stress from welding, cold working and other asymmetric metal treatments. In addition, water chemistry, welding, crevice geometry, heat treatment, and radiation can increase the susceptibility of a metal component to SCC. It is well known that SCC occurs at higher rates when oxygen is present in the reactor water in concentrations of about 1 to 5 parts per billion (ppb) or greater. SCC is further increased in components exposed to a high radiation flux where oxidizing species, such as oxygen, hydrogen peroxide, and short-lived radicals, are produced from radiolytic decomposition of the reactor cooling water. Such oxidizing species increase the electrochemical corrosion potential (ECP) of metals. Electrochemical corrosion is caused by a flow of electrons from anodic to cathodic areas on metallic surfaces. The ECP is a measure of the thermodynamic tendency for corrosion phenomena to occur, and is a fundamental parameter in determining rates of various corrosion phenomena, including SCC, corrosion fatigue, corrosion film thickening, and general corrosion. The ECP has been clearly shown to be a primary variable in controlling the susceptibility to SCC in BWR environments. FIG. 1 shows the observed (data points) and predicted (curves) crack growth rate as a function of corrosion potential for 25 mm CT specimens of furnace-sensitized Type 304 stainless steel at 27.5 to 30 MPa.sqroot.m constant load in 288.degree. C. water over the range of solution conductivities from 0.1 to 0.5 .mu.S/cm. Data points at elevated corrosion potentials and growth rates correspond to actual irradiated water chemistry conditions in test or commercial reactors. Corrosion (or mixed) potential represents a kinetic balance of various oxidation and reduction reactions on a metal surface placed in an electrolyte, and can be decreased by reducing the concentration of oxidants such as dissolved oxygen. FIG. 2 is a schematic of E (potential) vs. log .vertline.i.vertline. (absolute value of current density) curves showing the interaction of H.sub.2 and O.sub.2 on a catalytically active surface such as platinum or palladium. The terms i.sub.0 represents the exchange current densities, which are a measure of the reversibility of the reactions. Above i.sub.0, activation polarization (Tafel behavior) is shown in the sloped, linear regions. The terms i.sub.L represent the limited current densities for oxygen diffusion to the metal surface, which vary with mass transport rate (e.g., oxygen concentration, temperature, and convection). The corrosion potential in high-temperature water containing oxygen and hydrogen is usually controlled by the intersection of the O.sub.2 reduction curve (O.sub.2 +2H.sub.2 O+4e.sup.- .fwdarw.4OH.sup.-) with the H.sub.2 oxidation curve (H.sub.2 .fwdarw.2H.sup.+ +2e.sup.-), with the low kinetics of metal dissolution generally having only a small role. The fundamental importance of corrosion potential versus, for example, the dissolved oxygen concentration per se, is shown in FIG. 3, where the crack growth rate of a Pd-coated CT specimen drops dramatically once excess hydrogen conditions are achieved, despite the presence of a relatively high oxygen concentration. FIG. 3 is a plot of crack length vs. time for a Pd-coated CT specimen of sensitized Type 304 stainless steel showing accelerated crack growth at .apprxeq.0.1 .mu.M H.sub.2 SO.sub.4 in 288.degree. C. water containing about 400 ppb oxygen. Because the CT specimen was Pd-coated, the change to excess hydrogen caused the corrosion potential and crack growth rate to drop. In a BWR, the radiolysis of the primary water coolant in the reactor core causes the net decomposition of a small fraction of the water to the chemical products H.sub.2, H.sub.2 O.sub.2, O.sub.2 and other oxidizing and reducing radicals. For steady-state operating conditions, approximately equilibrium concentrations are established for O.sub.2, H.sub.2 O.sub.2, and H.sub.2 in the water which is recirculated and for O.sub.2 and H.sub.2 in the steam going to the turbine. The resultant concentrations of O.sub.2, H.sub.2 O.sub.2, and H.sub.2 produce an oxidizing environment and result in conditions that can promote intergranular stress corrosion cracking (IGSCC) of susceptible materials of construction. One well-known method employed to mitigate IGSCC of susceptible material is the application of hydrogen water chemistry (HWC), whereby the oxidizing nature of the BWR environment is modified to a more reducing condition. This effect is achieved by adding hydrogen gas to the reactor feedwater. When the hydrogen reaches the reactor vessel, it reacts with the radiolytically formed oxidizing species homogeneously and on metal surfaces to re-form water, thereby lowering the concentration of dissolved oxidizing species in the bulk water, including that portion of the water that is adjacent to metal surfaces. The rate of these recombination reactions is dependent on local radiation fields, water flow rates and other variables. In HWC, the injected hydrogen reduces the level of oxidizing species in the water, such as dissolved oxygen, and as a result lowers the ECP of metals in the water. However, factors such as variations in water flow rates and the time or intensity of exposure to neutron or gamma radiation result in different concentrations of the stated oxidizing species in different reactors, and different concentrations at different locations within the same reactor. Thus, varying amounts of hydrogen have been required to reduce the level of oxidizing species sufficiently to maintain the ECP below a critical potential required for protection of metal reactor components from IGSCC in high-temperature water. As used herein, the term "critical potential" means a corrosion potential at or below a range of values of about -0.230 to -0.300 V based on the standard hydrogen electrode (SHE) scale. IGSCC proceeds at an accelerated rate in systems in which the ECP is above the critical potential, and at a substantially lower rate, or effectively at a zero rate, in systems in which the ECP is below the critical potential (see FIG. 1 ). Water containing oxidizing species such as oxygen increases the ECP of metals exposed to the water above the critical potential, whereas water with little or no oxidizing species present results in an ECP below the critical potential. Initial use of HWC focused on relatively large additions of dissolved hydrogen, which proved capable of reducing the dissolved oxygen concentration in the water outside of the core from .apprxeq.200 ppb to <5 ppb, with a resulting change in corrosion potential from .apprxeq.+0.05 V.sub.SHE to .ltoreq.-0.25 V.sub.SHE. This approach is in commercial use in both domestic and foreign BWRs. Corrosion potentials of stainless steels and other structural materials in contact with reactor water containing oxidizing species can usually be reduced below the critical potential by the use of HWC through injection of hydrogen into the reactor feedwater. For adequate feedwater hydrogen addition rates, conditions necessary to inhibit IGSCC can be established in certain locations of the reactor. Different locations in the reactor system require different levels of hydrogen addition. Much higher hydrogen injection levels are necessary to reduce the ECP within the high radiation flux of the reactor core, or when oxidizing cationic impurities, for example, cupric ion, are present. It has been shown that IGSCC of Type 304 stainless steel (containing 18-20% Cr, 8-10.5% Ni and 2% Mn) and all other structural materials commonly used in BWRs can be mitigated effectively by reducing the ECP of the material to values below -0.230 V.sub.SHE. An effective method of achieving this objective is to use HWC. However, high hydrogen additions, for example, of about 200 ppb or greater in the water of the reactor core, that may be required to reduce the ECP below the critical potential, can result in a higher radiation level in the steam-driven turbine section from incorporation of the short-lived N.sup.16 species in the steam. For most BWRs, the amount of hydrogen addition required to provide mitigation of IGSCC of pressure vessel internal components results in an increase in the main steam line radiation monitor by a factor of five to eight. This increase in main steam line radiation can cause high, even unacceptable, environmental dose rates that can require expensive investments in shielding and radiation exposure control. Thus, recent investigations have focused on using minimum levels of hydrogen to achieve the benefits of HWC with minimum increase in the main steam radiation dose rates. In this context, it is important to recognize that significant mitigation of IGSCC can also occur when the corrosion potential is greater than -0.230 V.sub.SHE, such as when the corrosion potential is lowered by as little as 0.050 V.sub.SHE. Referring to FIG. 1, a reduction of 0.050 V.sub.SHE, for example, from -0.100 V.sub.SHE to -0.150 V.sub.SHE results in a reduction of the crack growth rate, at solution conductivities of 0.1-0.5 .mu.S/cm, by a factor of approximately two. Another effective approach used to reduce the ECP is to either coat or alloy the stainless steel surface with palladium or other noble metals. The presence of palladium on the stainless steel surface reduces the amount of hydrogen required to reach the required IGSCC critical potential of -0.230 V.sub.SHE. The use of alloys or metal coatings containing noble metals permits lower corrosion potentials (e.g., .apprxeq.-0.5 V.sub.SHE) to be achieved at much lower hydrogen addition rates. For example, U.S. Pat. No. 5,135,709 (Andresen et al.) discloses a method for lowering the ECP on components formed from carbon steel, alloy steel, stainless steel and other iron-base alloys, as well as nickel-base alloys or cobalt-base alloys which are exposed to high-temperature water by forming the component to have a catalytic layer of a noble metal. Such approaches rely on the very efficient recombination kinetics of dissolved oxygen and hydrogen on catalytic surfaces (see the high i.sub.0 for H.sub.2 oxidation in FIG. 2, which causes most O.sub.2 reduction curves to intersect at -0.5 V.sub.SHE). This was demonstrated not only for pure noble metals and coatings, but also for very dilute alloys or metal coatings containing, for example, <0.1 wt. % Pt or Pd (see FIGS. 3 to 5). FIG. 4 shows corrosion potential measurements on pure platinum, Type 304 stainless steel and Type 304 stainless steel thermally sprayed by the hyper-velocity oxy-fuel (HVOF) technique with a powder of Type 308L stainless steel containing 0.1 wt. % palladium. Data were obtained in 285.degree. C. water containing 200 ppb oxygen and varying amounts of hydrogen. The potential drops dramatically to its thermodynamic limit of .apprxeq.-0.5 V.sub.SHE once the hydrogen is near or above the stoichiometric value associated with recombination with oxygen to form water (2H.sub.2 +O.sub.2 .fwdarw.2H.sub.2 O). FIG. 5 shows corrosion potentials of Type 304 stainless steel doped with 0.35 wt. % palladium at a flow rate of 200 cc/min. in 288.degree. C. water containing up to 5000 ppb oxygen and various amounts of hydrogen. If the surface recombination rate is much higher than the rate of supply of oxidants to the metal surface (through the stagnant, near-surface boundary layer of water), then the concentration of oxidants (at the surface) becomes very low and the corrosion potential drops to its thermodynamic limit of .apprxeq.-0.5 V.sub.SHE in 288.degree. C. water, even though the bulk concentration of dissolved oxygen remains high (FIGS. 3 to 5). Further, the somewhat higher diffusion rate of dissolved hydrogen versus dissolved oxygen through the boundary layer of water permits somewhat substoichiometric bulk concentrations of hydrogen to support full recombination of the oxidant which arrives at the metal surface. While some hydrogen addition to BWRs will still be necessary with this approach, the addition can be vastly lower, as low as .ltoreq.1% of that required for the initial HWC concept. Hydrogen additions remain necessary since, while oxidants (primarily oxygen and hydrogen peroxide) and reductants (primarily hydrogen) are produced by radiolysis in stoichiometric balance, hydrogen preferentially partitions to the steam phase in a BWR. Also, no hydrogen peroxide goes into the steam. Thus, in BWR recirculation water there is some excess of oxygen relative to hydrogen, and then, in addition, a fairly large concentration of hydrogen peroxide (e.g., .apprxeq.200 ppb). Approaches designed to catalytically decompose the hydrogen peroxide before or during steam separation (above the core) have also been identified. While the noble metal approach works very well under many conditions, both laboratory data and in-core measurements on noble metals show that it is possible for the oxidant supply rate to the metal surface to approach and/or exceed the recombination rate (see FIGS. 6 and 7). FIG. 6 shows the effect of feedwater hydrogen addition on the corrosion potential of stainless steel and platinum at several locations at the Duane Arnold BWR located in Palo, Iowa. At .apprxeq.2 SCFM of feedwater hydrogen addition, the corrosion potentials in the recirculation piping drop below .apprxeq.-0.25 V.sub.SHE, However, in the high flux (top of core) regions, even for pure Pt, the corrosion potential remains above .apprxeq.-0.25 V.sub.SHE at feedwater hydrogen levels of .gtoreq.15 SCFM, where long-term operation is very unattractive due to the cost of hydrogen and the increase in volatile N.sup.16 (turbine shine). FIG. 7 shows corrosion potential vs. hydrogen addition for Pd-coated Type 316 stainless steel in 288.degree. C. water in a rotating cylinder specimen, which simulates high fluid flow rate conditions. The water contained 1.0 part per million (ppm) O.sub.2. As the hydrogen level was increased above stoichiometry, the potential decreased, but only to about -0.20 V.sub.SHE. The oxygen supply rate in these tests had exceeded the exchange current density (i.sub.0) of the hydrogen reaction (see FIG. 2), and activation polarization (Tafel response) of the hydrogen reaction began to occur, causing a shift to a mixed (or corrosion) potential which is in between the potentials measured in normal and extreme hydrogen water chemistry on non-catalytic surfaces. At the point where the oxidant supply rate to the metal surface approaches and/or exceeds the recombination rate, the corrosion potential will rapidly increase by several hundred millivolts (e.g., to .gtoreq.-0.2 V.sub.SHE). Indeed, even under (relatively small) excess hydrogen conditions, pure platinum electrodes in the core of BWRs exhibit corrosion potentials which are quite high, although still somewhat lower than (non-catalytic) stainless steel (see FIG. 6). At very high hydrogen levels (well above those typically used in the original hydrogen water chemistry concept), the corrosion potential on noble metal surfaces will drop to <-0.3 V.sub.SHE (see FIG. 6). However, the huge cost of the hydrogen additions combined with large observed increase in volatile radioactive nitrogen in the steam (i.e., N.sup.16, which can raise the radiation levels in the turbine building) make the use of very high hydrogen addition rates unpalatable. Therefore, it is desirable to develop other means for lowering the ECP of metal components in high temperature water in addition to HWC and catalytic coatings or alloys, particularly means that may overcome some or all of the limitations of these methods of lowering the ECP. SUMMARY OF THE INVENTION The present invention is an alternative method for achieving the objective of low ECPs which result in slow or no crack growth in stainless steel and other metals in high temperature water containing oxidant species such as hydrogen peroxide and dissolved oxygen. This is accomplished by coating the surfaces of IGSCC-susceptible reactor components with an electrically insulating material that is doped with a noble metal, such as zirconia. In accordance with the present invention, the metal corrosion potential is shifted in the negative direction without the addition of hydrogen. This invention may be briefly described as a method for mitigating growth of a crack in a surface of a metal component adapted for use in high-temperature water, an uncoated surface of the metal component being susceptible to stress corrosion cracking in high-temperature water, comprising the step of applying a coating on the surface of the metal component, the coating comprising an electrically insulating material that is doped with a noble metal, the coating having restricted mass transport crevices which penetrate to the surface of the metal component and which restrict the flow of oxidants to the surface, the electrically insulating material doped to a concentration of the noble metal that is sufficiently small to avoid the establishment of conductive paths through the coating from the surface of the metal component to an outer surface of the coating, the crevices also exposing the noble metal to the oxidants thereby promoting reduction of the oxidants with available reductants, whereby the corrosion potential of the surface of the metal component is decreased by at least 0.050 V by application of the coating. This invention is particularly advantageous in that these catalytic, insulating coatings lower the ECP both by limiting the transport of oxidant species to the surface of a metal component exposed to high temperature water and by the catalytic reduction of these species during the transport process. |
abstract | A method for extracting tritium from irradiated boiling water reactor control rods that have cruciform-shaped. Bands of a malleable metal are wrapped around the flat portions of the blades, one band near the top of each blade panel and a second band near the bottom. The bands are crimped and an inlet penetration is formed through one of the bands and the panel and an outlet penetration is formed through the second band and the panel. A termination of each end of a closed loop conduit is sealably connected to the inlet and outlet for transporting a carrier gas through the interior of the panel. The carrier gas passing through the interior transports the tritium out of the panel to a tritium getter filter to capture the tritium. The carrier gas then recirculates through the system. |
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description | The present invention relates to an apparatus for preparing a specimen. More particularly, it relates to an apparatus for isolating and obtaining a part of a sample of which state and shape may be changed with temperature change. It also relates to a sample-processing apparatus for preparing a specimen, and a method for evaluating the prepared specimen. With the increase of functional devices, demand for cross-sectional evaluation or fine processing of organic substances such as biological materials and plastics is increasing. Methods for making cross sections of organic materials in order to obtain structural information include cutting with a knife, resin embedding, freeze-embedding, freeze-fracture and ion etching. Usually, for internal structure observation of an organic matter by optical microscopy, the sample is embedded in a resin and sliced with a microtome. However, the optical microscope allows only macroscopic observation of the cross section and the cut-out position cannot be designated in this method, so that it requires an enormous amount of repeating work of preparing cross sections in order to observe and analyze the structure at a desired position. Recently, focused ion beam (FIB) techniques that can process a predetermined site have been developed, where a finely focused ion beam from an ion source irradiates a sample for processing it by etching etc. Such FIB etching has become considerably popular, and utilized widely for structural analysis or failure analysis of semiconductors or the like, and for sample preparation of scanning electron microscopy (SEM), transmission electron microscopy (TEM) etc. Recently, several methods have been proposed to cut out a portion of a sample and process it applying manipulation techniques to FIB techniques. For example, Japanese Patent Application Laid-Open No. H05-52721 proposes a method of cutting out a part of a sample by FIB and the cut out minute sample is held on a probe, which facilitates isolation of only the necessary portion for analysis. Japanese Patent Application Laid-Open No. 2001-345360 proposes a method of cutting out a minute specimen by using an ion beam and then bombarding it with another ion beam to reduce the influence of the element of the first ion beam employed for cutting out. However, when the sample is a material of which state and shape will change with temperature such as an organic substance, it is difficult to prepare a minute specimen of a desired-shape using such a probe since the temperature of such a probe often becomes higher than that of the sample resulting in heating of the contacted part of the sample. Thus, the present invention is to provide an apparatus for conveying a sample for observation without heat-denaturation. The present invention is also to provide an apparatus suitable for obtaining a necessary minute piece from a sample. The present invention is also to provide a sample processing apparatus capable of efficiently processing a necessary minute piece from a sample under temperature control of the sample. Furthermore, the present invention is to provide a sample evaluating apparatus and a sample evaluating method for analysis of a cross section structure under temperature control of the sample. Furthermore, the present invention is to provide a sample-conveying apparatus capable of conveying a sample for electron microscopic observation under temperature control of the sample. The sample-conveying apparatus of the present invention comprises a probe for conveying a specimen to be observed, and temperature control means for controlling a temperature of the probe whereby the sample does not change during conveyance. The specimen-obtaining apparatus of the present invention comprises a stage for supporting a sample; first temperature control means which controls a temperature of the sample; means for isolating a part of the sample; probe moving means for mounting and moving a probe; a probe for obtaining a part of the sample isolated by the isolation means; and second temperature control means for controlling a temperature of the probe. The sample-processing apparatus of the present invention comprises a stage for supporting a sample; first temperature control means for controlling a temperature of the sample; ion beam generation means for irradiating the sample with an ion beam; detection means for detecting a signal emitted from the sample in response to the irradiation of the ion beam; a probe for obtaining a part of the sample processed by the irradiation of the ion beam; a sample table for evaluation; second temperature control means for controlling a temperature of the probe; and third temperature control means for controlling a temperature of the sample table. The sample evaluation apparatus of the present invention is characterized in that ion beam irradiation is carried out by using ion beam generating means and information is acquired by the detection means with a sample preconditioned at the predetermined temperature by temperature-controlling means, the sample is cut out and pasted under the conditions the temperature of the probe and the sample has been adjusted at a predetermined temperature by temperature-controlling means. According to preferred embodiments, a sample table for evaluation may be provided separately from the stage. Also the temperature control means may be provided with cooling means which cools the sample to a temperature equal to or less than the room temperature. The stage, the ion beam generation means, the ion beam detection means, the probe and the sample table may be provided within a chamber with a controllable atmosphere, and there may be further provided trap means for trapping gas remaining in the chamber. Also the emission signal may be secondary electrons or secondary ions. Also the detection means may be constituted of a first detector for detecting secondary electrons and a second detector for detecting secondary ions. As described above, according to the present invention, temperature of the probe is regulated by the second temperature control means to maintain the temperature of the sample at a desired temperature, so that a minute specimen can be obtained from a sample of which state and shape are susceptible to temperature change. Also the first temperature control means allows to maintain the sample at a desired temperature during sample processing, for example, FIB operation. Furthermore, since the sample table for evaluation can be temperature-controlled by the third temperature control means, the sample after fixation to the sample table can be maintained at a desired temperature. Consequently, change in state or shape of the sample would not occur as in the prior technology. The sample processing method of the present invention comprises the steps of regulating temperature of a sample, a probe and a sample table; sectioning or processing the sample by irradiating a predetermined portion of the sample with an ion beam from at least two angular directions relative to a surface of the sample; and connecting the probe to a part of the sectioned sample. The sample-evaluation method of the present invention comprises the steps of regulating a temperature of a sample, a probe and a sample table; sectioning or processing the sample by irradiating a predetermined portion of the sample with an ion beam from at least two angular directions relative to a surface of the sample; connecting the probe to a part of the sectioned sample; isolating the sectioned sample to which the probe has been attached; attaching the isolated sample to the sample table using the probe; cutting off the probe; and irradiating the sample attached to the sample table with an evaluation beam for evaluation to obtain from an emitted signal an image of a cross-sectioned face of the sample generated by the sectioning or processing step. The present invention also provides a conveying apparatus which comprises a conveying member for conveying a sample for observation under an electron microscope; and temperature regulation means which regulates a temperature of the conveying member; wherein the temperature regulation means regulates the temperature of the sample in such a manner that it does not change before and after the conveyance. Now embodiments of the present invention will be explained with reference to accompanying drawings. FIG. 1 is a schematic view of a focused ion beam processing apparatus constituting a first embodiment of the sample-processing apparatus of the present invention. The processing apparatus is provided with a temperature holding unit 2 on which a sample 1 is fixed and which maintains the fixed sample 1 at a predetermined temperature. The temperature holding unit 2 can be accommodated in a sample chamber 3. The sample chamber 3 is provided with an ion beam generating unit 4 for irradiating the sample 1, fixed to the temperature holding unit 2, with an ion beam, an electron detector 5 for detecting secondary electrons emitted from the sample 1 by the ion beam irradiation, gas introducing means which enables a film deposition on the sample 1 by the ion beam irradiation, and a probe-holding unit 7 capable of mounting a probe (not shown) for fixing a part of the sample cut out by the ion beam irradiation. The probe-holding unit 7 is preferably a manipulator that serves as probe-moving means for moving the tip of the probe three-dimensionally. There is also provided a sample table 8 for facilitating evaluation with another analysis apparatus. The interior of the sample chamber 3 can be evacuated by a pump (not shown) and can be maintained at a predetermined low pressure, thereby enabling an ion beam irradiation. The interior of the sample chamber 3 is preferably maintained at a pressure within a range from 10−10 to 10−2 Pa. The ion beam generation unit 4 is used as means for isolating a part of the sample by irradiating the sample 1 with the ion beam, and also as processing means for processing the sample, for example, to reveal a cross-section of the sample. It can also be used for scanning ion microscopy (SIM) observation. In case of SIM observation, the ion beam generation unit 4 and the electron detector 5 serve as information acquisition means, and the secondary electrons generated when the sample 1 is irradiated with the ion beam are detected by the electron detector 5 and an image is formed on the basis of detection signals from the electron detector 5. The detection signal from the electron detector 5 is sent to a control unit 9 constituting control means for imaging, and imaging of the aforementioned SIM observation is executed by the control unit 9. For example, the control unit 9 receives image information (mapping information) in the detection signal from the electron detector 5 and displays the obtained image information as an image on an display apparatus (not shown). In addition, the control unit 9 controls ion beam generation in the ion beam generation unit 4 and controls irradiation and scanning of such ion beam onto the sample 1. The beam scanning can be controlled at the side of the unit 4, or at the side of the unit 2 on which the sample is fixed, or both, but control at the side of beam generating unit 4 is desirable in consideration of the scanning speed. Also the irradiating position of the ion beam can be so controlled as to focus it on the probe tip on the sample 1. The ion beam generation unit 4 may have such a configuration as described in Japanese Patent Application Laid-Open No. H05-52721 or No. 2000-217290. Probe The probe in the present embodiment is used for obtaining a sample piece (specimen) from a sample by fixing the piece to the probe tip (not shown), and the sample piece is separated from the sample by processing. Also the probe tip is preferably constituted of a material of a satisfactory thermal conductivity, in consideration of temperature control. Also the material preferably has a certain resistance to the ion beam, since SIM imaging with the FIB is used for confirming the position of the sample or the probe. Furthermore, the probe tip is consumed, cut off little by little after the specimen is attached to the sample table. For this reason, it is preferable that the probe or at least the probe tip can be replaced. Configuration of the First, Second and Third Temperature Control Means In the present embodiment, the first temperature control means is provided with a temperature holding unit 2 capable of temperature control of the sample. The temperature holding unit 2 is, for example, a sample stage provided with a temperature controller. FIG. 2 schematically shows a configuration of such a sample stage with a temperature controller. Referring to FIG. 2, the sample stage with the temperature controller is constituted of a sample stage 13 having a temperature-varying system 12 in a portion on which the sample 1 is fixed, a thermometer 11a for directly detecting the temperature of the sample 1, a thermometer 11b attached to a part of the temperature-varying system 12 to detect the temperature in the vicinity of the sample 1 fixed to the temperature-varying system 12, and a temperature control unit 9a for controlling the temperature of the temperature-varying system 12 on the basis of the temperature detected by the thermometer 11b, thereby maintaining the sample 1 at a predetermined temperature. Although not shown in FIG. 2, there is also provided a display unit for displaying the temperature detected by the thermometer 11b, and the operator can confirm the temperature of the sample 1 displayed on the display unit. Also the temperature control unit 9a may be so constructed as to regulate the temperature of the temperature varying system 12 according to the temperatures detected by both the thermometers 11a and 11b, and such configuration enables a more accurate temperature control of the sample 1. In the present embodiment, the sample table 8 for evaluation is also fixed to the temperature holding unit 2 like the sample 1, and can be controlled at a predetermined temperature. In such case the first temperature control means serves also as the third temperature control means, but such configuration is not restrictive and there may be provided separate temperature control means. The temperature varying system 12 is integrated with the thermometer 11b as a unit in the sample stage 13 to control the temperature in a necessary range. Such unit can be, for example, a high temperature unit having a heating mechanism such as a heater or a low temperature unit having a cooling mechanism. It may be, if necessary, a unit having a temperature varying function ranging from low to high temperature spanning room temperature. The sample stage 13 can mechanically move, rotate or incline the fixed sample 1 three dimensionally, and thus can move the sample 1 to a desired position for evaluation. The movement control of the sample 1 on the sample stage 13 is achieved by the aforementioned control unit 9. The aforementioned cooling mechanism can employ, for example, a Peltier element or a helium freezer. Otherwise, the aforementioned cooling mechanism may be constructed such that a cooling pipe for a coolant is provided in the temperature holding unit beneath the area where the sample is fixed so that a coolant such as liquefied nitrogen or water thermally comes in contact with the temperature holding unit. Also in order to increase the efficiency of heat absorption for heat generated during the processing operation, it is preferable to increase the contact efficiency between the sample and the cooling unit (temperature holding unit). Such a design can be realized, for example, by preparing a sample holder in such a configuration as to wrap up the sample but not to hinder optical systems employed in processing and in observation, or by processing the sample in such a shape as to match that of the holder so that the sample is held maintaining a maximum contact area. Furthermore, it is also possible to cover a non-worked area of the sample with a cooling member. In such case, the cooling member is to be so placed as not to intercept the beam. Also in this embodiment, a similar temperature holding unit may be incorporated in the probe as the second temperature control means for controlling the temperature of the probe. In such case, it may be controlled by another control unit, but it is also possible to use the control unit 9 for controlling the temperature of the sample and the probe. Also the temperature control means for controlling the temperature of the probe may be of such a configuration that the probe is connected to the probe moving means and they are in thermal contact thereby achieving the temperature control. Also the probe, the probe moving means and the second temperature control means may be constructed as shown in FIG. 7. Referring to FIG. 7, a probe 40 can be mounted on a probe holder 41 provided with a temperature varying system 12 having a thermometer 11. The probe holder 41 provided with the probe 40 is contained in the probe moving means. Also the probe holder 41 can move in the prove moving means 7 according to necessity for probe protection, and the tip of the probe 40 can retract in or jut from the probe moving means 7 according to the necessity. Also the temperature of the probe may be regulated by the second temperature control means, according to the temperature of the sample regulated by the first temperature control means. Also the first temperature control means and the second temperature control means preferably communicate electrically, and the temperature of the probe is synchronously regulated by the second temperature control means, according to the temperature of the sample regulated by the first temperature control means. The sample and the probe are preferably maintained at substantially the same temperature by the first and second temperature control means, but there may also be employed different temperatures. In such case, it is preferable to maintain the probe at a lower temperature. Cross Section Evaluating Method for Sample A cross section evaluation method of the present invention will be explained in the following. FIG. 3 is a flow chart showing a process of evaluation of a cross section of the sample, utilizing the sample processing apparatus shown in FIG. 1. In the following there will be given an explanation on the process of cross sectional evaluation with reference to FIG. 3, and also on the control for FIB processing and SIM observation by the control unit 9 and the temperature control of the sample by the temperature control unit 9a according to such process. First, the sample 1 and the sample table 8 are fixed on a predetermined position (temperature varying system 12) of the sample stage 13 (step S10), and, after an introduction thereof into the sample chamber 3, an evaluation temperature (temperature at which evaluation is carried out) is set (step S11). At the same time, the evaluation temperature of the probe is similarly set. Once the evaluation temperature is set, the temperature control unit 9a controls the temperature at the temperature varying system 12, whereby the sample 1 and the sample table 8 are maintained at such a set evaluation temperature. The temperature of the sample 1 is detected by the thermometer 11a, and the operator can confirm, by the detected temperature displayed on the display unit (not shown), whether the sample 1 is maintained at such evaluation temperature. In the present embodiment, it is preferred to cool the sample to a temperature lower than the room temperature for processing. Also by cooling to 0° C. or lower, it is possible to solidify water present in the sample. In case of employing such a cooling step, it is preferable to first cool the sample to the room temperature or lower, then to maintain the cooled sample in an atmosphere of a reduced pressure, and to execute the irradiation with the focused beam while absorbing the heat generated from the irradiated surface of the sample and the vicinity thereof, thereby maintaining the shape of the non-irradiated portion. Also in sample cooling, rapid cooling from the room temperature may be adopted. In such a case, preferably the cooling rate is 40° C./min or more, whereby. In this manner, it is possible to observe a cross section in a rapidly cooled state even when the sample is a mixture of which dispersibility varies by temperature. The cooling step is preferably executed before a pressure reducing step. In this manner it is possible to suppress evaporation of the sample under a reduced pressure. In case the sample is constituted of a substance of low evaporation, the cooling may be executed simultaneously with the pressure reduction. The probe may be cooled after the pressure reduction. The cooling step varies depending on the sample to be processed, but, in case of an ordinary organic substance such as PET, cooling is preferably executed within a temperature range of 0 to −200° C., more preferably −50 to −100° C. If the processing time or the cooling time becomes too long and the sample stays in a low temperature state too long, a gas remaining in the sample chamber or a substance generated in the processing may be adsorbed on the cooled sample, rendering the desired processing or observation difficult. For this reason, it is preferable to provide trap means for adsorbing the remaining gas or the substance generated at the processing, and to execute the processing and the information acquisition while cooling such trap means. The present invention is advantageously applicable to a case where the sample to be processed is a substance susceptible to heat, such as an organic substance, particularly a protein or another biological substance, or a composition containing water. It is particularly preferable for a composition containing water, since the sample can be worked in a state retaining water. Since the irradiation of the focused ion beam is executed under a reduced pressure, FIB processing of a composition containing water or highly volatile organic molecules may cause evaporation of water by the heat generated during the processing. Therefore the temperature control means provided in the present invention is highly effective. For achieving processing and structural evaluation more precisely, it is also preferable to carry out a step of determining in advance the suitable holding temperature of processing. In such a case, a sample equivalent to the sample to be worked is employed as a reference, and is worked at various temperatures, and preferred holding temperature is determined by investigating the relation between the damage in the worked portion and the cooling temperature. After the sample 1 is confirmed to be maintained at the evaluation temperature, the surface of the sample 1 is subjected to SIM observation under constant monitoring of the temperature of the sample 1 (step S12). In such SIM observation, there is employed a weak ion beam for observation, and the sample 1 is scanned by the ion beam from the ion beam generation unit 4 under the control by the control unit 9 on the ion beam irradiation from the ion beam generation unit 4 and on the movement of the sample stage 13. Also in synchronization with such scanning, the electron detector 5 detects the secondary electrons, and the control unit 9 displays a SIM image on the display unit (not shown), based on a detection signal of such secondary electrons. In this manner, the operator can observe the SIM image of the surface of the sample 1. Then a cross section evaluating position (position subjected to cross-section evaluation) is precisely determined from the image obtained by the SIM observation of the surface of the sample 1 (SIM image displayed on the aforementioned display unit) (step S13), and thus determined cross section evaluating position is further subjected to a SIM observation with a processing beam (step S14). In such SIM observation, the sample 1 is scanned within the range of the cross section evaluating position by the ion beam from the ion beam generation unit 4 under the control by the control unit 9 on the ion beam irradiation from the ion beam generation unit 4 and on the movement of the sample stage 13. Also in synchronization with such scanning, the electron detector 5 detects the secondary electrons, and the control unit 9 displays a SIM image on the display unit (not shown), based on a detection signal of such secondary electrons. In this manner, the operator can observe the surface SIM image of the cross section evaluating position determined in the step S14. Also after the determination of the cross section evaluating position and before the SIM observation with the processing beam, if necessary, a gas is introduced from the gas introducing means 6 for depositing a protective film on the periphery of the sample 1 including a worked portion. Then FIB processing conditions are set (step S15). In such setting of the FIB processing conditions, the area and position to be sectioned is determined on the SIM image obtained in the surface SIM observation of the step S14, and there are also set cross section processing conditions including an accelerating voltage, a beam current and a beam diameter. The cross section processing conditions include crude processing conditions and fine processing conditions, both of which are set at this state. The crude processing conditions have a beam diameter and an energy amount larger than those in the fine processing conditions. The area and position to be sectioned may be determined on the SIM obtained with the observing ion beam in the step S12, but, in consideration of precision, it is preferable to determine on the SIM image utilizing the ion beam to be employed in the actual processing. In such processing from the surface, an ion beam is irradiated in an amount necessary for cutting onto the area and position to be sectioned determined in step S15, where the control unit 9 controls the ion beam generation unit 4 to the preset processing conditions and also controls the movement of the sample stage 13. In this operation, in order to isolate a part of the sample, including the portion necessary for evaluation, most of the periphery of the cut-out sample, seen from above, is worked deeper than the portion for evaluation, only leaving a portion which can be cut off by a minimum processing thereafter. After the processing from the surface, the surface of the sample 1 is subjected to a SIM observation, and there is confirmed, on the image obtained by the SIM observation (SIM image), whether the processing has proceeded close to the desired position (step S17). Then the sample is inclined together with the stage (step S18), and a cross section prepared by the processing from the surface is subjected to a SIM observation with an observing beam, in order to confirm the shape of the surface (step S18). In case the processing has not proceeded close to the desired depth, the stage is returned to the original inclination and the foregoing steps S16 to S19 are repeated. Then FIB processing conditions from the direction of cross section are set (step S20). In such setting of the FIB processing conditions, a sectioned area and a sectioned position are determined on the SIM image obtained in the cross sectional SIM observation of the step S19, and there are also set cross section processing conditions including an accelerating voltage, a beam current and a beam diameter. Then there is executed an FIB processing from the cross sectional direction (step S21). The cross sectional direction means an angular direction different from the surface of the sample, and allowing observation of the cross section formed by the FIB processing from the sample surface. Therefore, it need not be perpendicular to the cross section surface. In such processing from the cross sectional direction, most parts are isolated from the sample, only leaving the aforementioned portion which can be cut off by a minimum processing. Then, after the sample is inclined to the original angle (step S22), a weak beam for observation is irradiated from the surface, thereby executing a surface SIM observation of the sample 1 (step S23). Based on the surface SIM image in the step S23, the probe is moved onto a minute sample piece to be removed (step S24). Then it is confirmed that the probe is contacted with the minute piece for example by a contrast change in the SIM image, a gas is introduced by the gas introducing means 6 and an FIB beam is irradiated in a position including the contact portion between the probe and the minute piece to deposit a film, thereby adhering the probe (step S25). Then the aforementioned portion which can be cut off by a minimum processing is cut off with a processing beam (step S26) whereby the sample piece is isolated from the sample 1 and remains in a state fixed to the probe. In this state, since the sample including the sample piece and the probe are maintained at equivalent temperatures by the step S11, the temperature of the sample piece scarcely changes at least by the contact of the probe. If the temperature of the probe and the sample is significantly different, the temperature of the sectioned sample piece will change close to the temperature of the probe. Thus, by confirming, prior to the contact of the probe, that the temperature of the sample and the probe is almost the same, it is possible to prevent the temperature change in the sample piece. Thereafter, the probe with the sample piece adhered thereto is moved onto the evaluating sample table 8 (step S27), then, after confirmation that the sample piece is in contact with the sample table 8, a gas is introduced from the gas introducing means 6 and an FIB beam irradiation is executed to adhere the sample piece to the sample table 8 (step S28). Thereafter a part of the probe is cutoff by an ion beam (step S29), whereby the sample piece is transferred to the sample table 8. Also, if necessary, the sample stage 13 is inclined to cause an inclination in the sample table 8, and a film deposition is executed by a gas introduction and an ion beam irradiation thereby reinforcing the adhesion between the sample piece and the sample table 8. Also in order to facilitate the adhesion and the cutting-off of the probe, the tip of the probe is not in a position perpendicular to the sample surface nor to the cross section of the sample but is inclined by a certain angle, thereby enabling a positional confirmation by the SIM image and a processing with the ion beam. After the sample piece is adhered to the sample table 8 for evaluation as explained above, a SIM observation of the sample surface is executed by an observing beam (step S30) to confirm the adhesion in a desired position, and subsequently an FIB processing (finish processing) is executed (step S31). In the finish processing, the sample piece fixed to the sample table 8 by the process up to the step S29 is irradiated with an ion beam of an amount necessary for the finish processing, under the control by the control unit 9 on the ion beam generation unit 4 with the aforementioned finish processing conditions and on the movement of the sample stage 13. Such finish processing allows preparation of a smooth cross section suitable, for example, for an observation of a high magnification under a transmission electron microscope. Finally a SIM observation of the sample surface (step S32) is executed, in order to obtain an evaluation sample of a desired thickness. In case the processing has not been made to the desired thickness, the aforementioned steps S31 and S32 are repeated. Also a cross section for the scanning electron microscope can be prepared by a finish processing of one surface. As explained in the foregoing, the sample processing apparatus of the present embodiment can constantly maintain the sample 1 to be evaluated, the probe and the sample table 8 for evaluation at a set temperature, so that the sample 1 is not changed in the state or the shape thereof. Consequently even a sample easily damaged by processing can be exactly evaluated for a fine structure. The cross section processing method in the foregoing embodiments is effective for analyzing, at a desired temperature, samples on various substrates such as glass including a polymer structure, microparticles, a polymer structure containing liquid crystal, a particle dispersion structure in fibrous materials or a material causing a temperature transition. It is also effective for a sample easily damaged by an ion beam. Also the aforementioned probe moving means may be provided in plural units in the apparatus. In the present embodiment, in addition to the configuration of the embodiment 1, there is provided, as shown in FIG. 4, trap means 14 for preventing re-deposition of a residual gas in the sample chamber 3 or a substance generated in the processing onto the sample 1. The trap means 14 is constituted, for example, of a metal of a satisfactory thermal conductivity, and is maintained, during the cooling of the sample 1, at a temperature equal to or lower than that of the sample 1. The present embodiment provides an effect of preventing deposition of an impurity onto the sample 1, in case the sample 1 is worked and observed in a state maintained equal to or lower than the room temperature. For example, in case of the above-explained FIB-assisted deposition, there may result a layer of impurity between the deposition layer and the worked sample, whereby a desired function is hindered, but the present embodiment suppresses formation of such an impurity layer by the trap means 14. The trap means 14 is provided, in a state where the sample stage supporting the sample 1, the ion beam generation unit 4, the electron detector 5, the gas introducing means 6 and the probe are positioned, in such a position as not hindering the detection system and the beam system at the processing. The trap means 14 is preferably provided in a position not hindering such detection and processing and as close as possible to the sample 1, in order to improve the trapping efficiency. Also the trap means 14 may be provided in one or more positions in the sample chamber 3 maintained at a low pressure. In the embodiment, there will be explained a case of utilizing the apparatus of the present invention as a cross section processing apparatus in the production process of a liquid crystal display apparatus or an organic semiconductor. More specifically, there will be explained an embodiment of executing a temperature control of a sample of a relatively large area. In case of exactly evaluating a cross sectional state in a part of a large-sized sample such as a glass substrate coated with liquid crystal, for use in a large-sized liquid crystal display apparatus, the temperature control may be executed either in a local area in the vicinity of a worked portion, or over the entire substrate. In case of temperature control of the entire substrate, a coolant pipe for passing a coolant is provided in a position of the temperature holding unit under the surface supporting the sample, thereby cooling the entire holder. In the following, there will be explained actual examples of cross sectional evaluation of samples with the sample evaluating apparatus of the foregoing embodiments. In this example, there was employed a sample-processing apparatus for cross sectional observation, shown in FIG. 1. The temperature holding unit 2 was constituted of the sample stage with the temperature controller as shown in FIG. 2, incorporating a unit with a low temperature-varying system, and a cross sectional evaluation of a sample, bearing a polymer structure (polymerizable monomers HEMA, R167, HDDA polymerized together with liquid crystal) containing liquid crystal (two-frequency drivable liquid crystal DF01XX manufactured by Chisso Inc.) on a glass substrate, was executed in the following process. The sample was fixed with a carbon paste to a unit with a low temperature-varying system, and the unit was set on the sample stage 13. The sample stage 13 with the sample was inserted in the sample chamber 3, which was then evacuated to a predetermined low pressure. Then the evaluation temperature was set at −100° C., and it was confirmed that the sample was maintained at such evaluation temperature. A surface SIM observation was executed on an area including a cross sectional observation portion of the sample, under constant monitoring of the sample temperature. Based on an image obtained by the surface SIM observation, an approximately central portion of the sample was determined as a cross sectional observation portion. Then the determined cross sectional observation portion was irradiated with an ion beam to acquire a SIM image. The ion beam in this operation was a very weak one of the observation mode. More specifically, there were employed a gallium ion source, an acceleration voltage of 30 kV, a beam current of 20 pA and a beam diameter of about 30 nm. A cross section processing position was designated on the acquired SIM image. Then the designated cross section processing position was subjected to an FIB processing (surface processing). More specifically, there were employed an acceleration voltage of 30 kV, a beam current of 50 pA and a beam diameter of 300 nm to form, in the cross sectional processing position, a rectangular recess of a square shape of 40 μm and a depth of 30 μm. Then an L-shaped recess was formed similarly with a processing beam so as to be connected with the rectangular recess, leaving an evaluation portion. Then the sample was inclined, and, after a confirmation of the processing to the desired position with a weak beam for SIM observation, the bottom portion of the remaining evaluation portion was subjected to an FIB processing (cross section processing). The inclining angle was about 60°. FIG. 5A is a schematic view of the sample prepared by such FIB processing. At an approximate center of a sample 30, there were formed, by the irradiation of the ion beam 20 and by the following ion beam irradiation with the inclination of the sample, a rectangular recess, an L-shaped recess connected to the rectangular recess and a slat-shaped sample piece of which bottom is detached by an ion beam 21 after the inclination of the sample and which is partially connected to the sample. Then the inclined sample was returned to the original state, then the probe was contacted with the sample piece partially connected to the sample, and a film was formed in the contact position of the probe and the sample piece by introducing a gas from the gas introducing means and executing an FIB beam irradiation, thereby adhering the probe to the sample piece. Thereafter the partial connecting portion between the sample piece and the sample was FIB processed to cut off the sample piece from the sample, and the probe was elevated to lift the sample piece together with the probe. FIG. 5B is a schematic view showing thus sectioned sample piece. The sample piece partially connected to the sample 30 as shown in FIG. 5A was adhered to the probe 40, then separated from the sample 30 by cutting off the connecting portion, and was lifted as a sample piece 31. In this operation, it was confirmed that the probe was maintained at a temperature of about −100° C., the same as that of the sample. For film formation, there was employed a gas of tungsten carbonyl (W(CO)6). Then the sample piece on the probe was moved, together with the probe, onto the sample table, contacted with the sample table under temperature control and adhered to the sample table by a gas introduction as in the case of probe adhesion, and the probe used for moving the sample piece was cut off by an FIB processing. FIG. 5C is a schematic view showing the sample table on which the sample piece was adhered. The sample piece 31 was fixed by the sample table 50 by a deposition film 60. Then a finish processing was conducted. In this case, a thin piece was prepared for TEM observation. In such finish processing, the processing was conducted with gradually weaker conditions, and, during the processing, the sample surface under processing was SIM observed from time to time for confirming whether the processing proceeded close to the desired position. Finally, for improving the precision of cross sectional processing, the cross sectional processing portion was further worked under a weak condition similar to that in the SIM observation. In the present example, as explained in the foregoing, the FIB processing was conducted while the sample was maintained at −100° C., so that the cross sectional processing could be executed without deforming of the liquid crystal layer during the processing. After a thin slice for TEM was prepared in this manner, the sample was returned to the normal temperature and was TEM observed, whereby a cross section of the polymer layer structure on the substrate could be observed. In the present example, a cross sectional evaluation of polymer particles (polystyrene) prepared on a PET substrate was executed in the following procedure. The apparatus shown in FIG. 1 was set at a temperature of about 10° C., and the sample and the probe were temperature controlled. After the tip of the probe was contacted with the polymer particle to be evaluated as in Example 1, and the polymer particle and the probe were adhered by a gas introduction and an FIB irradiation. FIG. 6A is a schematic view showing the probe 40 and the sample 31 thus adhered, and a deposition film 60. Then the polymer particle was moved together with the probe, and FIB processing was executed at the cross sectional observation portion. FIG. 6B is a schematic view showing the probe 40 and the sample 31 after the cross sectional processing. After cross sectional processing of the polymer particle in this manner, the sample was taken out together with the probe and was subjected to a cross sectional SEM observation and an elemental analysis in another evaluation apparatus. At first the SEM observation proved that the polymer particle had a uniform interior without a bubble. There were employed conditions of an acceleration voltage of 15 kV and a magnification up to about 30,000. Then, during the SEM observation, specific X-ray emitted from the cross section of the sample 31 was acquired to prepare a mapped image (elemental analysis), thereby proving that aluminum was dispersed in the polymer. In the foregoing, there has been explained a method of evaluating a cross section of a sample, but the present invention is not limited thereto. For example, the present invention includes a configuration where a substance deposited to the surface is removed, thereby exposing the surface to be observed and executing a surface observation. |
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claims | 1. A fuel element storage and cooling configuration, comprising:a fuel element storage pool;a heat sink constructed as a natural draft dry cooling tower disposed at a horizontal distance of between 20 meters and 100 meters from said fuel element storage pool;a cooling system including:at least one first heat exchanger configured as an evaporator, disposed in said fuel element storage pool and having a highest point, andat least one second heat exchanger configured as a condenser and disposed in said heat sink at a distance from said at least one first heat exchanger of 3 m-5 m above said highest point of said at least one first heat exchanger; anda pipe system at least partially filled with a flowable coolant and interconnecting said at least one first and said at least one second heat exchangers to form a closed circuit, said pipe system having a part forming a return line with a continuous slope for cooled or condensed coolant, and said pipe system ensuring natural circulation of the coolant and thus heat transport from said fuel element storage pool to said heat sink, without a pump apparatus, upon a temperature increase of said at least one first heat exchanger relative to said at least one second heat exchanger. 2. The fuel element storage and cooling configuration according to claim 1, wherein the flowable coolant is a refrigerant. 3. The fuel element storage and cooling configuration according to claim 1, wherein said at least one first heat exchanger is suspended. 4. The fuel element storage and cooling configuration according to claim 1, which further comprises a baffle disposed on an outer surface of said at least one first heat exchanger. 5. The fuel element storage and cooling configuration according to claim 1, which further comprises closure devices provided in a point-by-point manner inside said closed circuit, said closure devices being activated in the event of a sudden pressure drop to stop circulation of the coolant. 6. The fuel element storage and cooling configuration according to claim 1, wherein said cooling system is one of a plurality of cooling systems being independent of one another at least with respect to said closed circuit. 7. The fuel element storage and cooling configuration according to claim 6, wherein said cooling systems are configured in an overredundant manner. 8. The fuel element storage and cooling configuration according to claim 6, wherein said cooling systems are configured in a diverse manner. 9. The fuel element storage and cooling configuration according to claim 1, wherein said cooling system is configured for a cooling output of 5 MW-30 MW. |
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048636746 | abstract | A nuclear reactor installation with a simple configuration includes a transport container which may be placed on the cover of the cavity. The transport container bottom overlaps the cover access opening. A vertical displaceable pebble conveyor line extends from the pebble pile surface to the transport container interior. A blower is located between the transport container and the reactor. The blower suction line is connected to the transport container interior and the blower pressure line leads into the reactor vessel. |
051911574 | summary | BACKGROUND OF THE INVENTION 1. FIELD OF THE INVENTION This invention relates to a method and apparatus for permanent disposal of hazardous waste and more particularly, to a method and apparatus for permanent disposal of hazardous waste using a borehole extending through a geopressured formation. 2. CROSS REFERENCE TO RELATED APPLICATION The disclosure of this application is related to my prior application having the same title, now all abandoned, as follows: ______________________________________ Ser. No. Filing Date ______________________________________ 06/468,842 June 20, 1983 06/621,518 June 18, 1984 07/018,757 February 24, 1987 07 147/040 January 20, 1988 ______________________________________ 3. DESCRIPTION OF THE PRIOR ART Permanent disposal of hazardous waste, such as flammables, heavy metals, acids and bases and synthetic organic chemicals present difficult problems. The difficulties are especially acute in disposal of heavy metals (radioactive) waste. Many methods have been developed to provide a proper and safe disposal of hazardous waste without contaminating our natural resources. Various disposal methods include landfills, injection wells, incineration, ocean dumping, wastes exchange, and destruction through organisms ("superbug" method). U.S. Pat. No. 4,335,978 to Mutch discloses a land fill disposal system. Rather than relying upon the subsurface formation itself to prevent fluid migration, a pair of spaced impermeable liners are employed to prevent fluid migration. The disclosed land fill is located above rather than below the subsurface water table for the area to preclude contamination. Haynes et al U.S. Pat. No. 4,377,509 is entitled "Packaging for Ocean Disposal of Low-Level Radioactive Waste Material". A plurality of conventional 55 gallon metal drums are filled with the nuclear waste material and placed within a concrete shell. A filler material of asphaltic or a dry portland cement concrete is then used to fill the shell. Immediately prior to dumping in the ocean, water is introduced into the shell to activate the cement. In an alternate embodiment, the concrete is allowed to harden before dumping into the ocean. U.S. Pat. No. 4,377,167 to Bird et al discloses two improved container materials for solid waste materials at an underground impervious stable rock formation. The prior practice had been to rely upon the insolubility of the radioactive elements to prevent migration of the radioactive waste material rather than containing the waste for a sufficient period of time to effect decay within the container. Bird's invention resides in forming a container out of a naturally occurring nickel alloy having proven superior aging characteristics. The Upermann U.S. Pat. No. 4,316,814 is entitled "Seal For A Storage Borehole Accommodating Radioactive Waste and Method of Applying the Seal". The storage waste containers are lowered into the borehole formed in a salt formation in a stacked relationship. The seal of the borehole above the stored material prevents escape of the radioactive waste up the borehole. The Klingle et al U.S. Pat. No. 4,252,462 is entitled "Chemical Land Fill" for disposal of waste water sludge. An impoundment area having a liquid impervious base and a perimeter dike is arranged to receive the waste water liquid therein. The sludge is dewatered and subsequently covered with an impervious layer. The following patents to Gablin et al disclose systems for disposing of nuclear reactor effluent having mixed liquid and particulate matters: U.S. Pat. Nos. 4,196,169, 4,168,243, 4,056,362, 4,167,491, 3,986,977. Geologists have characterized subsurface rock formations forming the earth's crust in various ways. One such classification has been to divide sedimentary rocks into two broad groups based on their pore-fluid pressures. These two mutually exclusive groups are labeled (1) hydropressures and (2) geopressures, and will be defined in this application as such. Hydropressure zones or formation have pore fluid pressures that are created by the effective weight of the overlying waters plus the back pressure of out-flowing waters. Geopressure formations or zones are created where the hydropressure rock is sealed in a confined geological container (geopressure cell) and is subjected to a geostatic pressuring source greater than hydropressures. The geostatic pressuring force source is the weight and temperature of the earth's crust with depth of burial. A classic example of a hydropressure-geopressure province is the Gulf of Mexico Salt Basin, which includes the Texas-Louisiana Cenzoic Salt Basin. Hydropressure formations have leaks which enable flow or migration of the fluid pressure so over time they adjust to the hydropressure pressure for the depth. This is commonly referred to as normal pressure. Unlike hydropressures or hydropressure formations, geopressure formations are sealed. A geopressure seal is defined as a restriction to flow such that geopressures have not been dissipated between the time they were created in the geologic past and the present. By definition all geopressures or geopressure formations must have a geopressure seal. The block of the earth's crust that is sealed off and contains the geopressures is called a geopressure cell which is the definition adopted herein. To create a geopressure cell (a confined or enclosed container or reservoir), the surrounding earth crust formations must be effective as a seal at the top, bottom and all sides of the cell. The geopressure cells or formations are sealed in regional fault blocks by shale layers and regional fault growths. Porosity is preserved in geopressure formation or zones due to the pore fluid pressure which is greater than the hydropressure for the same depth. They are sometimes called or referred to as abnormally high-pressure zones or formations in the petroleum industry. For an in depth description of hydropressure geopressure formations and their characteristics and properties, see the article "Geopressures" by Charles A. Stuart which appears in the Supplemental Proceedings of the Second Symposium on Abnormal Subsurface Pressure presented Jan. 30, 1970 at Louisiana State University in Baton Rouge, La. The encountering of geopressure zones when drilling for hydrocarbons or minerals presented substantial problems. In U.S. Pat. No. 3,399,723, to Charles A. Stuart (class 166 subclass 4) those drilling problems associated with encountering a geopressure formation are addressed, but not for the purpose of the present invention. From the standpoint of describing the present invention both hydropressure and geopressure formations are defined and explained at length in the Stuart patent. The problem of encountering the abnormally high pressure of the geopressure zone when the geopressure barrier seal (the transition or mutation zone) is broken or penetrated by the drill bit is described as a kick and the parameters of accommodating that pressure transition are addressed. All of the above specifically mentioned or identified U.S. patents and the C. A. Stuart published article are hereby fully and specifically incorporated herein for forming part of applicant's written description as if their content had been set forth in full. IDENTIFICATION OF THE OBJECTS OF THE INVENTION An object of the present invention is to provide a method for disposing hazardous waste in a borehole extending into a subsurface geopressured formation or cell. It is another object of this invention to provide a method that will provide for permanent safe disposal of radioactive and/or other hazardous waste in boreholes formed in geopressure formations. Another object of this invention is to provide a disposal method where sealed containers filled with hazardous waste are encased within a subsurface geopressured formation to prevent waste migration in the event of container failure. A further object of the present invention is to provide a method of disposal of hazardous waste by hydraulically injecting hazardous waste into the pores of a geopressured formation or cell to trap the waste within the geopressured formation. SUMMARY OF THE INVENTION The present invention relates to a method for permanently disposing hazardous waste in a borehole extending through a sealed, non-migrating geopressured formation. In one embodiment the waste is sealed in an elongated container which is lowered down the borehole to concentrically position the containers in a stacked relationship in the borehole within a geopressured formation. The stacked containers are then completely encased within the borehole for restoring the geopressure formation seal. A second embodiment of the method for disposing hazardous waste in a borehole extending through a non-migrating geopressured formation includes pumping the dissolved or entrained hazardous waste down the borehole and injecting the waste through perforations in the casing into the pores of the geopressured formation. The pressurized hazardous waste may fracture the geopressured formation and upon reduction in the pressure the waste is trapped in the geopressured formation which is then resealed to restore its geopressure characteristics. |
054250706 | description | DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS Referring to FIG. 1, there is shown a perspective view of a nuclear fuel assembly insert alignment tool 10 in accordance with the present invention. The insert alignment tool 10 includes a frame having an upper mounting bracket 12 and a lower fixed plate 24 which are coupled together to form a rigid structure by means of first and second support pipes 20a and 20b. Support pipes 20a, 20b are on the order of 25 feet in length to accommodate the length of the rodlets in a typical nuclear fuel assembly insert and maintain sufficient water cover for radiation shielding. The insert alignment tool 10 is shown suspended from the wall 32 of a spent fuel pool in the side elevation view of FIG. 2. As stated, the typical length A of the insert alignment tool 10 is on the order of 25 feet, with the distance B in a typical spent fuel pool being on the order of 15 feet. Attached to a lower surface of the tool's mounting bracket 12 is a positioning bracket 14 which is adapted for engaging an outer surface of the spent fuel pool's wall 32 for securely maintaining the insert alignment tool 10 and a nuclear fuel assembly with inserts (not shown) with which it is employed securely in position in a spent fuel pool. Positioning bracket 14 is affixed to the lower surface of mounting bracket 12 by conventional means such as weldments. Insert alignment tool 10 further includes an elongated, cylindrical shaft, or crank, 16 extending through mounting bracket 12 and attached at a lower end thereof to the fixed plate 24. Rotating shaft 16 includes a handle 18 at the upper end thereof and is provided with appropriate bearings 19 and 21 in the mounting bracket 12 and fixed plate 24 to facilitate its rotational displacement in response to a manual movement by a worker turning the handle. The lower end of rotating shaft 16 is pivotally coupled to the fixed plate 24 by bearing 21. FIG. 3 is perspective view of an upper portion of the insert alignment tool 10 including its mounting bracket 12 illustrating the manner in which a support member 34 (shown in dotted line form) such as a high strength line may be wrapped around the first and second support pipes 20a and 20b for lifting the insert alignment tool. Referring to FIG. 4, there is shown a perspective view of the lower portion of the insert alignment tool 10 including its fixed plate 24. FIGS. 5 and 6 are top plan views of the insert alignment tool's fixed plate 24 showing a movable plate 25 respectively in the full open and full closed positions. Drive linkage 30 connects the lower end of rotating shaft 16 to movable plate 25 and includes a pivoting arm 30a and a coupling bracket 30c pivotally connected together by means of a coupling pin 30b. The proximal end of pivoting arm 30a is attached to the lower end of the rotating shaft 16 by means of weldment 30e, while the proximal end of coupling bracket 30c is connected to the movable plate 25 by means of a coupling pin 30d. Movable plate 25 is disposed on the upper surface of fixed plate 24 and is maintained in position thereon by means of a first alignment bracket 36 engaging a first lateral edge of the movable plate and second and third alignment brackets 38 and 40 engaging a second, opposed edge of the movable plate. Each of the first, second, and third alignment brackets 36, 38 and 40 is securely attached to the upper surface of the fixed plate 24 by means of screws or bolts and allow the movable plate 25 to slide on the fixed plate's upper surface. When shaft 16 is rotated in a clockwise direction as shown by the direction of arrow 70 in FIG. 5, the movable plate 25 is displaced in the direction of arrow 72 to the fully retracted position as shown in the figure. When the movable plate 25 is moved leftward in the direction of arrow 72 to the fully retracted position, an edge of the plate engages a pair of retraction stop brackets 56 mounted to the upper surface of the fixed plate 24. When shaft 16 is rotated in a counterclockwise direction as shown by the direction of arrow 74 in FIG. 6, the movable plate 25 is displaced in the direction of arrow 76 to the closed position as shown in the figure. Continued counterclockwise rotation of shaft 16 results in pivoting arm 30a of drive linkage 30 engaging an extension stop bracket 50 mounted to the upper surface of the fixed plate 24. This limits rightward displacement of the movable plate 25 and ensures that it is in the fully closed position as shown in FIG. 6. As shown in FIGS. 4, 5 and 6, movable plate 25 includes a first comb structure 26 having a plurality of elongated, linear, pointed teeth 26a with a space, or gap, 26b disposed between adjacent teeth. Similarly, fixed plate 24 includes a second comb structure 28 also comprised of a plurality of teeth 28a separated by spaces 28b. The inner set of teeth within each of the first and second comb structures 26, 28 are longer and wider than the outer teeth. Fixed plate 24 further includes a pair of bevelled edges 78a and 78b disposed adjacent the second comb structure 28, while movable plate 25 similarly includes a pair of bevelled edges 80a and 80b disposed adjacent the first comb structure 26. Disposed in movable plate 25 are a pair of first locking apertures 42a and 44a, while disposed in fixed plate 24 are a pair of second locking apertures 42b and 44b. With movable plate in the fully closed position as shown in FIG. 6, first and second locking pins 46 and 48 may be inserted through aligned apertures 42a, 42b and 44a, 44b, respectively. With locking pins 46, 48 inserted through aligned pairs of apertures in the fixed and movable plates 24, 25, the rodlets 52 are maintained in fixed relative position in a locked manner. The nuclear fuel assembly insert alignment tool 10 is used with a nuclear fuel assembly insert having a plurality of spaced rodlets suspended therefrom in the following manner. A matrix array of rodlets 52 of a nuclear fuel assembly insert are first positioned adjacent the alignment tool's fixed plate 24. Fixed plate 24 is then urged into engagement with the rodlets 52 so that linear arrays of rodlets are received into each of the spaces 28b between adjacent teeth 28a in the second comb structure 28. The rodlets 52 are arranged in a general matrix array as shown in FIG. 5 which illustrates the second comb structure 28 as fully enclosing the array of rodlets. The pointed ends of teeth 28a as well as the first and second bevelled edges 78a and 78b leading into the second comb structure 28 facilitate positioning of the rodlets within the second comb structure even when the rodlets are entangled and misaligned. The two innermost teeth in the comb structure 28 are longer than the outer teeth which permits the center row of rodlets 52 to be initially aligned and received by the first comb structure followed by alignment between the shorter teeth of the outer rows of rodlets. Next, the movable plate 25 is displaced leftward in the direction of arrow 76 by means of the rotating shaft 16 so that the rodlets 52 are further positioned in the spaces 26b between adjacent teeth 26a in the first comb structure 26. This is shown in FIG. 6, where the movable plate 25 is illustrated in the fully closed position so as to completely overlap the second comb structure 28 in fixed plate 24. In this configuration, each of the rodlets 52 is securely maintained in fixed relative position by the respective teeth in the first and second comb structures 26, 28. The rodlets 52 are prevented from becoming misaligned and entangled by this overlapping comb arrangement and the space occupied by the rodlets is minimized to facilitate their insertion into the guide tubes of a nuclear fuel assembly for their storage in a spent fuel pool. Referring to FIG. 7, there is shown a top plan view of a burnable poison rod assembly 60 with which the nuclear fuel assembly insert alignment tool of the present invention is intended for use. The engagement of the rodlets 68 of the burnable poison rod assembly 60 by the nuclear fuel assembly insert alignment tool 10 of the present invention is shown in the side elevation view of FIG. 8, where the burnable poison rod assembly is shown in dotted line form. The burnable poison rod assembly 60 includes an upper frame 64 from which is suspended by means of a connecting shaft 66 a spoked attachment hub 62. Each of the rodlets 68 is securely suspended from the attachment hub 62 by means of a respective connector 58. As shown in the figure, the insert alignment tool's fixed plate 24 is disposed adjacent the upper end of the rodlets 68 for preparing to maintain the rodlets in alignment. In FIG. 8, the movable plate 25 is shown in the fully retracted position prior to movement to the closed position for engaging the rodlets 68. The lower part of the rodlets are aligned by raising the insert after closing the movable plate. There has thus been shown a nuclear fuel assembly insert alignment tool for engaging and aligning the rodlets, or pins, of a nuclear fuel assembly insert and maintaining the rodlets in fixed relative position for storage in a spent fuel pool via insertion into the guide tubes of a nuclear fuel assembly. The insert alignment tool allows a single worker to align the rodlets in a fixed, matrix array and is adapted for use with virtually any conventional nuclear fuel assembly insert such as a Rod Cluster Control Assembly, Burnable Poison Rod Assembly, or a thimble plug. The insert alignment tool is of simple construction, affords reliable operation, and can be used with even badly bowed rodlets in the nuclear fuel assembly. If a rodlet is damaged or bowed beyond use, the tool can still be used after removing the unusable rodlet with an underwater cutting tool. While particular embodiments of the present invention have been shown and described, it will be obvious to those skilled in the art that changes and modifications may be made without departing from the invention in its broader aspects. Therefore, the aim in the appended claims is to cover all such changes and modifications as fall within the true spirit and scope of the invention. The matter set forth in the foregoing description and accompanying drawings is offered by way of illustration only and not as a limitation. The actual scope of the invention is intended to be defined in the following claims when viewed in their proper perspective based on the prior art. |
summary | ||
051320770 | description | DETAILED DESCRIPTION OF THE ILLUSTRATED EMBODIMENT The number 10 generally designates the improved lower end fitting constructed according to the invention. The end fitting 10 is attached by means of instrumentation tube 12 and guide tubes 14 to the remaining fuel assembly structure, as represented by lower grid 16 and fuel rods 18 partially shown in FIG. 1. A typical pressurized water nuclear reactor lower core plate 20 is shown in dotted lines in FIG. 1. Diffusing flow between the multi-holed lower core plate 20 and the multi-holed top flow plate 22 of the lower end fitting, below the fuel rods 18, are the bell mouth type flow diffusers 30. As flow in the reactor core leaves the openings 24 in lower core plate 20 it is directed to and enters the lower end and smaller ends 32 of diffusers 30. As seen in FIGS. 3 and 4, the curved shape from smaller end 32 to larger output end 34 of each diffuser 30 is a parabolic type adjusted in their bell mouth shape in a manner dictated by the particular lower core plate flow hole geometry, lower end fitting design, and the reaction flow rate to achieve maximum jet diffusion. This, in turn, produces a resultant reduction in damaging vibration of fuel rods 18 and pressure drop through turbulence. The diffusers also act to stiffen the lower end fitting 10 by being rigidly connected to the legs 36 at the corners of the lower end fitting by horizontal webs 38 which are integral with legs 36. The diffusers 30 are mounted in recesses in webs 38 defined by counterbore 40. Also adding rigidity to the lower end fitting is a wear-reduction-shield 50 mounted in a central opening 52 in web 38. The opening 52 and web 38 form a hub which provides the central opening for an in-core-instrumentation thimble (not shown). This structure will be clearly understood from reading my U.S. Pat. No. 4,888,149. It will be apparent from the drawings here presented as FIGS. 2-4, that wear-reduction-shield 50 will, when the proper relationship between lower end fitting 10 and the core plate 20 are established by pins 56 and holes 58 in web 38, act further to stiffen the end fitting 10. Moreover, the relationship of shield 50 to a counterbore 60 in top flow plate 22 and counterbore 52 is such that it will help prevent tilting or tipping of the end fitting 10 and therefore prevent legs 36 from entering holes 24 in core plate 20. Again, see my U.S. Pat. No. 4,888,149. |
description | The present invention relates generally to an electromagnetic energy spot curing system, and more particularly to a system that utilizes an external output signal from an external processor to control the operation of the UV curing unit. It is well known to use ultraviolet (UV) lamps to cure certain curable compounds such as adhesives and the like. UV spot curing systems are used in various applications including the curing of industrial sealants for potting electronics, bonding plastics in the medical industry and the curing of dental filling materials, disk industry amongst other applications. Commercially available UV spot curing assemblies typically include a UV light or visible source, a reflector by which reflected light from the light source is focused on a target location of an object. Currently, spot curing systems use an internal timing device to control the exposure time of the light on the object. One known light curing system is disclosed in U.S. Pat. No. 5,803,729. The light cure system described therein includes a light source, a light guide for delivering the light produced by the source to a work site, a dimmer for controlling the intensity of light delivered to the work site, and a shutter for controlling the exposure time for the work site. The system also includes a controller for controlling both the dimmer and the shutter by adjusting the exposure time and/or intensity level so that a predetermined quantity of light energy is delivered to the work site. As it is apparent, the UV light curing system shown in the '729 patent utilizes an internal timer, which is commercially available, to control the shutter actuation. A trip signal is usually a ground connection that is made by use of an external footswitch. To interface with an external processor a relay must be used to complete the trip circuit. In the electromechanical relay, it is advantageous to eliminate oxidation and/or wear of the contacts in order to avoid either partial or total loss of current. Also, a timer must be pre-set independent of the main processing unit. This arrangement is costly due to the additional cost associated with a timer integrated to the curing unit and makes it less flexible to be interfaced with an external system or processor. It is therefore desirable to provide a UV curing lamp assembly wherein the assembly may be more economically and easily manufactured, and further, allowing the operation of the assembly unit to be externally controlled and its function is fully integrated with other operational components of the complete automated system. The present invention provides an electromagnetic energy spot curing system and method of achieving the same. The system includes a curing unit having a radiation source positioned to irradiate a work piece with radiation energy; and a shutter for moving selectively to allow exposure of radiation energy from the source. The system further includes a processor external to the curing unit for providing an external output signal to control the exposure of radiation energy wherein the processor is integrated with the curing unit. The subject invention relates to an electromagnetic energy spot curing system which utilizes a source of radiation found in the electromagnetic spectrum (e.g., ultraviolet (UV); visible light (VIS); infrared). To describe the invention and illustrate its functioning, reference is made herein to the use of a UV lamp. It is to be understood that the UV lamp can be interchanged with other sources of electromagnetic energy. In addition, the electromagnetic energy source may provide electromagnetic energy of varying intensities and/or of varying wavelengths (e.g., various types of radiation). With reference to FIGS. 1 and 2, a UV energy spot curing system is generally shown and designated therein with the reference numeral 10. The system 10 includes a UV lamp 12 mounted between a lamp retainer 11 and a lamp mounting plate 13. The UV lamp 12 preferably is supplied within an elliptic reflector 14 so that the energy emanated from the UV lamp 12 is focused at a precise desired location such as an adhesive on a work-piece surface. A shutter 15 shaped as an inverted “L” is positioned generally parallel to the lamp mounting plate 13 to selectively control UV energy emanating from the lamp 12. The shutter 15 includes shutter guides 15a and 15b placed on top and bottom of the shutter 15 and an opening of generally circular configuration (not shown) located preferably between the guides. The shutter guides 15a and 15b basically place the shutter 15 in place. As will be discussed later, the shutter 15 is moveable to slide up and down within the guides 15a and 15b. The system 10 further includes a light guide receptacle 16 which is a hollow tube positioned opposite the shutter 15 from the UV lamp 12 with a focused UV energy beam E being directed into an inlet aperture 17 of the light guide receptacle 16. A light guide 18 is positioned to be supported by the light guide receptacle 16. Light guide receptacle 16 is a hollow tube having an opening 16a for supportable receipt of the light guide 18 therein. Light guide receptacle 16 is formed of aluminum, although other materials may be used. The light guide 18 may be secured in the light guide receptacle 16 by screws (not shown) affixed thereto through threaded openings 16b. The light guide 18 may be glass, optical fiber or any other suitable light transmission material known in the art, and is preferably of the liquid-filled type. The light guide 18 includes a light entrance end 18a which plugs into the receptacle 16 and a light output end 18b which may be directed to a work site or target which contains adhesive material to be light cured. The light guide receptacle 16 acts to at least partially collimate the UV beam E, as the beam is emitted from the light guide receptacle 16 into the light guide 18 and exposed to the work site via light output end 18b. The UV lamp 12 may preferably be a conventional straight mercury arc lamp, metal halide mercury lamp, a xenon-metal halide lamp or any other suitable radiation emitting lamps known in the art. The lamp 12 is controlled by a ballast 23. Ballast 23 is a known electrical device or chip used in fluorescent and HID fixtures for starting and regulating fluorescent and high intensity discharge lamps. Ballast 23 acts as a power regulating source providing sufficient increasing power for the lamp 12 and further controlling the level of power supplied to the lamp 12. To ensure alignment of the UV lamp 12, the shutter 15, the lamp mounting plate 13 and the light guide 16, it is preferred that all these components be fixed to a common base plate 112 to minimize misalignment therebetween. A rear support bracket 114 may be fastened to the base plate 112 to rigidify the structure. The mechanism for movement of the shutter 15 is described herein. The shutter 15 is generally planar as shown in FIGS. 1 and 2 is connected to one end of a mounting plate (not shown). The mounting plate has a hole through which a light guide receptacle 16 passes through. The shutter 15 is positioned generally planar to the lamp 12 to selectively control UV energy emanating from the system 10. Preferably, the shutter 15 is located to selectively control UV energy entering the inlet aperture 17. The projecting bottom portion 15c of the shutter is suitably connected to a solenoid 19 in the common base plate 112. The opening and closing of the shutter 15 is controlled by a solenoid 19 which may preferably be coupled to the shutter 15. Upon receiving appropriate signals, the solenoid 19 is activated or energized to push the shutter 15 to move up and down the guides 15a and 15b. This movement of the shutter 15 will either allow or prevent the transmission of light through the light guide 16. In other words, when solenoid 19 is activated, it moves the shutter 15 in a first position wherein the circular hole of the shutter 15 coincides or aligns with the light guide 18, thereby allowing the energy/light beam E from the UV lamp 12 to pass through the opening of shutter 15 into the inlet aperture 17 to the light entrance end 18a and emit light via the light output end 18b to the worksite. In a second position the opening of the shutter 15 is not aligned with the light guide 18 which covers the inlet aperture 17 thereby preventing the light beam E from the lamp 12 from reaching the light entrance end 18b. Preferably, the shutter selectively controls UV energy entering the inlet aperture 17. Although not shown, a rotating template may preferably be placed in alignment with a UV lamp 12 to vary UV energy intensity. The appropriate signals needed to activate the solenoid 19 are received from a PLC 20 or other known processor external to the UV spot curing assembly 10 in accordance with the present invention. As shown in FIG. 2, PLC 20 provides analog voltage signals to a transistor in the curing unit. The preferred transistor type is a metal-oxide-semiconductor-field-effect-transistor (MOSFET) 21. MOSFET is a device commercially available and is used in power electronics applications to amplify electrical signals. The MOSFET typically provides both current and voltage gain yielding an output current into an external load which exceeds the input current and an output voltage across that external load which in turn exceeds the input voltage. In the present invention, the external load is the solenoid 19 of FIGS. 1 and 2. The analog output signal preferably a low voltage signal ranging from 5 V to 25 V of the external PLC 20 provides voltage to gate of MOSFET 21, thereby turning on the MOSFET 21. The MOSFET 21, then, as mentioned earlier, yields an output current, which runs up to the coil of the solenoid 19, thereby activating the solenoid 19 to move the shutter 15, allowing transmission of light through the light guide 16. In this manner, the exposure of the light energy and its exposure time is controlled externally by the processor 20. Furthermore, in an alternate embodiment, the curing unit can be operated independent from PLC 20. An override switch 22 as shown in FIG. 2 is preferably connected to the MOSFET 21 to have an option to manually open and close the shutter 15 from an external foot switch 24. The switch 22 of FIG. 2 is shown in a position where the MOSFET 21 is prevented from operating, breaking the source of the MOSFET 21 to ground. There will be no current running in MOSFET 21 to activate the solenoid 19. However, when the switch 22 is turned upward it is connecting the MOSFET 21 to the ballast 23. The ballast 23 then supplies the power to turn on the MOSFET 21 which in turn activates the solenoid 19. PLC 20 may or may not be providing the voltage signal to the MOSFET 21. In this position, the voltage signal is being provided by ballast 23 which is part of the curing unit 10. In either manner, the MOSFET 21 will continue to receive the voltage signal to continue the operation of the curing unit 10. The overall operation of the curing unit 10 as discussed above is controlled from the external processor PLC 20. In other words, the connection of the PLC 20 to the gate of MOSFET 21 turns the MOSFET 21 on which in turn activates the solenoid 19 to control the shutter 15 actuation. This overall operation or process is of the curing unit 10 is completely automated and fully integrated with the operational components of the PLC 20. As a result of this operation, material can be cured with exposure and the dwell periods of the radiation energy being controlled externally. The devices and the system of the present invention can be used in conjunction with a variety of different photocurable adhesive compositions. For example, UV curable vinyl and (meth)acrylate-containing compositions, which may also be optionally anaerobically curable, may be employed. Such compositions may include urethane-acrylate copolymers and block copolymers such as those disclosed in U.S. Pat. Nos. 3,425,988; 4,295,909; and 4,309,526. Other useful photocurable compositions containing reactive (meth)acrylate components are disclosed in U.S. Pat. Nos. 4,415,604; 4,424,252; and 4,451,523, all to Loctite Corporation. Photoinitiators which are intended to be active primarily in the ultraviolet (UV) region are incorporated along with the curable component, and which upon exposure to sufficient ultraviolet light initiate photopolymerization of the curable component. Such UV compositions can be used as structural adhesives, potting compounds, gap filling compounds, sealing compounds, conformal coatings as well as other applications known to those skilled in the art. In addition to the aforementioned adhesive compositions, UV curable silicone compositions are also contemplated as being useful with the present invention. Such compositions contain a curable silicone component and a UV photoinitiator component. Additionally, cyanoacrylate adhesives designed to cure upon exposure to photoirradiation may also be employed. Examples of commercially available UV curing compositions include Loctite product numbers Adhesive 352, 3321, 3491, 3525 and 3201. Having described the preferred embodiments herein, it should be further appreciated that various modifications may be made thereto without departing from the contemplated scope of the invention. As such, the preferred embodiments described herein are intended in an illustrative rather than a limiting sense. The true scope of the invention is set forth in the claims appended hereto. |
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summary | ||
abstract | A device that will enable material to be irradiated as needed to produce a desired transmutation product inside the core of a nuclear reactor. The device provides a means for monitoring neutron flux in the vicinity of the material being irradiated to allow determination of the amount of transmutation product being produced. The device enables the irradiated material to be inserted into the reactor and held in place at desired axial positions and to be withdrawn from the reactor when desired without shutting down the reactor. The majority of the device may be re-used for subsequent irradiations. The device also enables the simple and rapid attachment of unirradiated target material to the portion of the device that transmits the motive force to insert and withdraw the target material into and out of the reactor and the rapid detachment of the irradiated material from the device for processing. |
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043572987 | abstract | Advantage is taken of the non-uniform axial neutron flux density distribution in a nuclear reactor core by using fuel rod spacers of low neutron absorption in high neutron flux density regions and fuel rod spacers of low coolant flow resistance in the lower neutron flux density regions of the core, this spacer combination also providing higher fuel bundle thermal limits. |
abstract | A system for storing nuclear fuel assemblies includes a plurality of cells housed within a support structure. A first cell may house a first fuel assembly and a second cell may house a second fuel assembly. A plurality of compartments separate the plurality of cells and provide passageways for coolant entering a bottom end of the support structure to remove heat from the nuclear fuel assemblies. A first perforation transfers coolant between the first cell and one or more of the compartments, and a second perforation transfers coolant between the second cell and the one or more compartments. At least a portion of the coolant entering the bottom end of the support structure is transferred between the plurality of cells and the plurality of compartments. Two or more fuel storage racks may be stacked together in alternating fuel patterns to facilitate cooling the fuel assemblies with liquid or air. |
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